Probabilistic Fracture Mechanics Analyses Comparison to LBB Assessments

2019 ◽  
Author(s):  
Robert Kurth ◽  
Cedric Sallaberry ◽  
Elizabeth Kurth ◽  
Frederick Brust
Keyword(s):  
Probabilistic Models ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Margins ◽  
First Results ◽  
Primary Water ◽  
Reactor Loop ◽  
The Individual ◽  
Leak Before Break ◽  
Metal Weld

Abstract Analysis of a generic dissimilar metal weld (DMW) susceptible to primary water stress corrosion cracking (PWSCC) in a pressurized water reactor (PWR) is used to compare the newly developed probabilistic models (xLPR code) to the previously performed deterministic leak before break (LBB) analyses. The objective of this scoping analysis is to develop a generic reactor loop composed of representative welds and to investigate the safety margins in the presence of PWSCC at the Alloy 82/182 locations. These locations have been previously studied and approved for LBB, however not in the presence of active degradation such as PWSCC. The purpose of this study is to investigate potential increase in risk due to this mechanism. Comparisons of the individual weld probabilistic results to the deterministic LBB analysis are made as the first results of this study. Additionally the individual welds are combined into a configuration representative of the primary loop. This configuration is then tested against the criterion recommended by the xLPR acceptance group. This xLPR criterion is then compared to the existing LBB criterion to assess the change, if any, in risk due to PWSCC.

Rupture Hardware Minimization in Pressurized Water Reactor Piping

1989 ◽  
Vol 111 (1) ◽  
pp. 64-71 ◽  
Author(s):  
S. K. Mukherjee ◽  
J. J. Szy Slow Ski ◽  
V. Chexal ◽  
D. M. Norris ◽  
N. A. Goldstein ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Ferritic Steel ◽  
High Energy ◽  
Pressurized Water Reactor ◽  
Pipe Failure ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Water Reactor ◽  
Leak Before Break ◽  
Generic Sense

For much of the high-energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station—Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.

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.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Crack Growth Behavior of Nickel Alloy Welds in a PWR Environment

2007 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Crack Growth ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Crack Growth Behavior ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Ni Alloy ◽  
The Individual

A program is under way at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated Light Water Reactor (LWR) coolant environments. This paper focuses on the cracking behavior of Ni-alloy welds in simulated pressurized water reactor (PWR) environment at 290–350°C. Crack growth tests have been conducted on both field- and laboratory-produced welds. The results are compared with the existing crack-growth-rate (CGR) data for Ni-alloy welds to determine the relative susceptibility of specific Ni-alloy welds to environmentally enhanced cracking. To analyze the CGRs, a superposition model was used to establish the individual contributions of mechanical fatigue, corrosion fatigue, and stress corrosion cracking.

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.

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