Study the Cyclic Plasticity Behavior of 508 LAS under Constant, Variable and Grid-Load-Following Loading Cycles for Fatigue Evaluation of PWR Components

Technical Report · Thu Sep 01 00:00:00 EDT 2016

DOI:https://doi.org/10.2172/1331616· OSTI ID:1331616

Mohanty, Subhasish ^[1]; Barua, Bipul ^[1]; Soppet, William K. ^[1]; Majumdar, Saurin ^[1]; Natesan, Ken ^[1]

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1. Argonne National Lab. (ANL), Argonne, IL (United States)

This report provides an update of an earlier assessment of environmentally assisted fatigue for components in light water reactors. This report is a deliverable in September 2016 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2016 report, we presented a detailed thermal-mechanical stress analysis model for simulating the stress-strain state of a reactor pressure vessel and its nozzles under grid-load-following conditions. In this report, we provide stress-controlled fatigue test data for 508 LAS base metal alloy under different loading amplitudes (constant, variable, and random grid-load-following) and environmental conditions (in air or pressurized water reactor coolant water at 300°C). Also presented is a cyclic plasticity-based analytical model that can simultaneously capture the amplitude and time dependency of the component behavior under fatigue loading. Results related to both amplitude-dependent and amplitude-independent parameters are presented. The validation results for the analytical/mechanistic model are discussed. This report provides guidance for estimating time-dependent, amplitude-independent parameters related to material behavior under different service conditions. The developed mechanistic models and the reported material parameters can be used to conduct more accurate fatigue and ratcheting evaluation of reactor components.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE), Reactor Fleet and Advanced Reactor Development. Nuclear Reactor Technologies
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 1331616
• Report Number(s): ANL/LWRS-16/03; 130694; TRN: US1700506
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PWR TYPE REACTORS
FATIGUE
PLASTICITY
DYNAMIC LOADS
AMPLITUDES
AIR
EVALUATION
STRESSES
TIME DEPENDENCE
RATCHETING
VALIDATION
REACTOR MATERIALS
STRESS ANALYSIS
ALLOYS
MATHEMATICAL MODELS
COOLANTS
RANDOMNESS
TEMPERATURE RANGE 0400-1000 K
Chaboche model
Grid load following fatigue cycles
PWR environmental fatigue
Variable amplitude fatigue test
cyclic plasticity