skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Assessment of the Efficiency of HWCon IASCC Crack Growth Rate for High Fluence BWRMaterials

Abstract

This report describes the experimental study performed to assess the efficiency of hydrogen water chemistry on the propagation rate of cracks generated by irradiation assisted stress corrosion cracking in high fluence material. The selection of the material and the test procedures followed for this study are presented. The test results obtained with 8.6 dpa specimen are discussed.

Authors:
 [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1369371
Report Number(s):
INL/EXT-16-39983
TRN: US1703382
DOE Contract Number:
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; CRACKING; STRESS CORROSION; HYDROGEN; CRACKS; CRACK PROPAGATION; WATER CHEMISTRY; IASCC; Irradiation assisted stress corrosin cracking

Citation Formats

Teysseyre, Sebastien Paul. Assessment of the Efficiency of HWCon IASCC Crack Growth Rate for High Fluence BWRMaterials. United States: N. p., 2016. Web. doi:10.2172/1369371.
Teysseyre, Sebastien Paul. Assessment of the Efficiency of HWCon IASCC Crack Growth Rate for High Fluence BWRMaterials. United States. doi:10.2172/1369371.
Teysseyre, Sebastien Paul. 2016. "Assessment of the Efficiency of HWCon IASCC Crack Growth Rate for High Fluence BWRMaterials". United States. doi:10.2172/1369371. https://www.osti.gov/servlets/purl/1369371.
@article{osti_1369371,
title = {Assessment of the Efficiency of HWCon IASCC Crack Growth Rate for High Fluence BWRMaterials},
author = {Teysseyre, Sebastien Paul},
abstractNote = {This report describes the experimental study performed to assess the efficiency of hydrogen water chemistry on the propagation rate of cracks generated by irradiation assisted stress corrosion cracking in high fluence material. The selection of the material and the test procedures followed for this study are presented. The test results obtained with 8.6 dpa specimen are discussed.},
doi = {10.2172/1369371},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

Save / Share:
  • As nuclear power plants age, the increasing neutron fluence experienced by stainless steels components affects the materials resistance to stress corrosion cracking and fracture toughness. The purpose of this report is to identify any new issues that are expected to rise as boiling water reactor power plants reach the end of their initial life and to propose a path forward to study such issues. It has been identified that the efficiency of hydrogen water chemistry mitigation technology may decrease as fluence increases for high-stress intensity factors. This report summarizes the data available to support this hypothesis and describes a programmore » plan to determine the efficiency of hydrogen water chemistry as a function of the stress intensity factor applied and fluence. This program plan includes acquisition of irradiated materials, generation of material via irradiation in a test reactor, and description of the test plan. This plan offers three approaches, each with an estimated timetable and budget.« less
  • Irradiation assisted stress corrosion cracking affects the structural reliability of austenitic stainless steels exposed to high flux in BWRs. The focus of this investigation was to measure the susceptibility of low fluence types 304 and 316L stainless steels to irradiation assisted stress corrosion cracking (IASCC) and to evaluate the contribution of instantaneous radiation effects on IASCC susceptibility. IASCC susceptibility has been extensively investigated by out-of-flux tests, primarily by the slow strain rate (SSR) technique, which only include effects of accumulated radiation damage. Results from comparative in-flux tests would reveal the applicability of out-of-flux tests data for predicting in-core materials` behavior.more » In conclusion, the IASCC susceptibility determined for CP type 304 and type 316L by in-flux SSR tests was similar to that found in out-of-flux SSR tests conducted at high oxygen concentrations. Mild susceptibility to IASCC was observed in the pre-irradiated CP 304 alloy and no IASCC was observed on the pre-irradiated 316L specimens. For low fluence CP type 304 and type 316L, the IASCC susceptibility, fracture mode and dependence on mechanical, microchemical and electrochemical parameters evaluated by in-flux SSRT were comparable to results determined by out-of-flux SSRT. This indicated that in-flux and out-of-flux environments with the same ECP are equivalent.« less
  • Grain boundary chromium carbides improve the resistance of nickel based alloys to primary water stress corrosion cracking (PWSCC). However, in weld heat affected zones (HAZ's), thermal cycles from fusion welding can solutionize beneficial grain boundary carbides, produce locally high residual stresses and strains, and promote PWSCC. The present research investigates the crack growth rate of an A600 HAZ as a function of test temperature. The A600 HAZ was fabricated by building up a gas-tungsten-arc-weld deposit of EN82H filler metal onto a mill-annealed A600 plate. Fracture mechanics based, stress corrosion crack growth rate testing was performed in high purity water betweenmore » 600 F and 680 F at an initial stress intensity factor of 40 ksi {radical}in and at a constant electrochemical potential. The HAZ samples exhibited significant SCC, entirely within the HAZ at all temperatures tested. While the HAZ samples showed the same temperature dependence for SCC as the base material (HAZ: 29.8 {+-} 11.2{sub 95%} kcal/mol vs A600 Base: 35.3 {+-} 2.58{sub 95%} kcal/mol), the crack growth rates were {approx} 30X faster than the A600 base material tested at the same conditions. The increased crack growth rates of the HAZ is attributed to fewer intergranular chromium rich carbides and to increased plastic strain in the HAZ as compared to the unaffected base material.« less
  • Evaluation of a crack-opening-displacement (COD) technique for measurement of crack length in compact tension (CT) specimens for the determination of fatigue crack growth rate (FCGR) has proved it to be successful. Comparisons of COD-determined FCGR data with previously established trend lines for two high-strength steels (5 Ni-Cr-Mo and 10Ni) are presented together with data from aluminum alloy 2024-T351. (auth)
  • The goal of the Mechanisms of Irradiation Assisted Stress Corrosion Cracking (IASCC) task in the LWRS Program is to conduct experimental research into understanding how multiple variables influence the crack initiation and crack growth in materials subjected to stress under corrosive conditions. This includes understanding the influences of alloy composition, radiation condition, water chemistry and metallurgical starting condition (i.e., previous cold work or heat treatments and the resulting microstructure) has on the behavior of materials. Testing involves crack initiation and growth testing on irradiated specimens of single-variable alloys in simulated Light Water Reactor (LWR) environments, tensile testing, hardness testing, microstructuralmore » and microchemical analysis, and detailed efforts to characterize localized deformation. Combined, these single-variable experiments will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long-term, design IASCC-resistant materials. In support of this research, efforts are currently underway to arrange shipment of “free” high fluence austenitic alloys available through Électricité de France (EDF) for post irradiation testing at the Oak Ridge National Laboratory (ORNL) and IASCC testing at the University of Michigan. These high fluence materials range in damage values from 45 to 125 displacements per atom (dpa). The samples identified for transport to the United States, which include nine, no-cost, 304, 308 and 316 tensile bars, were relocated from the Research Institute of Atomic Reactors (RIAR) in Dimitrovgrad, Ulyanovsk Oblast, Russia, and received at the Halden Reactor in Halden, Norway, on August 23, 2016. ORNL has been notified that a significant amount of work is required to prepare the samples for further shipment to Oak Ridge, Tennessee. The preliminary work for sample shipment between Halden and Oak Ridge includes fabrication of an inner cask sample container, decontamination and preparation of a Type A container, preparation of new activity calculations, all necessary paperwork, and handling. ORNL will continue to work to track progress of sample preparation and shipment status, and to work toward an agreement that covers material shipping costs between the Halden Reactor and the Oak Ridge National Laboratory.« less