scholarly journals Bounding Surface Flaw Configuration Susceptible to Stress Corrosion Cracking Under Welding Residual Stress in a Multiple-Purpose Canister

2017 ◽  
Author(s):  
Poh-Sang Lam ◽  
Robert L. Sindelar ◽  
Andrew J. Duncan ◽  
Joe T. Carter
Keyword(s):  
Residual Stress ◽  
Stress Intensity ◽  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Management Program ◽  
Corrosion Cracking ◽  
Surface Flaw ◽  
Welding Residual Stress

The part-through-wall crack perpendicular to the circumferential weld on the outside surface of a spent nuclear fuel (SNF) multiple-purpose canister (MPC) can be shown to be the most limiting fracture configuration driven by the welding residual stress (WRS). A series of semi-elliptical cracks of various sizes is chosen to calculate the stress intensity factors (K) under a bounding residual stress (i.e., the stress distribution that bounds all WRS in a canister). The threshold stress intensity factor (KISCC) of the canister material in the storage environment is used to determine a critical flaw size, below which the stress corrosion cracking would be unlikely to take place. This result can be considered as the flaw disposition criterion should a surface flaw be detected during the inservice inspection as required by the aging management program (AMP), and can be proposed to American Society of Mechanical Engineers (ASME) Section XI Code Case N-860, “Examination Requirements and Acceptance Standards for Spent Nuclear Fuel Storage and Transportation Containment Systems.”

NRC/EPRI Residual Stress Validation Program Phase I: Experimental Specimen Modeling and Measurement

2011 ◽  
Author(s):  
J. Broussard ◽  
P. Crooker
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Phase 1 ◽  
Measurement Techniques ◽  
Corrosion Cracking ◽  
Welding Residual Stress ◽  
Hole Drilling ◽  
Pressurized Water

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI weld residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper describes the Phase 1 program, which comprised an initial period of learning and research for both FEA methods and measurement techniques using simple welded specimens. The Phase 1 specimens include a number of plate and cylinder geometries, each designed to provide a controlled configuration for maximum repeatability of measurements and modeling. A spectrum of surface and through-wall residual stress measurement techniques have been explored using the Phase 1 specimens, including incremental hole drilling, ring-core, and x-ray diffraction for surface stresses and neutron diffraction, deep-hole drilling, and contour method for through-wall stresses. The measured residual stresses are compared to the predicted stress results from a number of researchers employing a variety of modeling techniques. Comparisons between the various measurement techniques and among the modeling results have allowed for greater insight into the impact of various parameters on predicted versus measured residual stress. This paper will also discuss the technical challenges and lessons learned as part of the DM weld materials residual stress measurements.

scholarly journals Preventing Stress Corrosion Cracking of Spent Nuclear Fuel Dry Storage Canisters

2019 ◽  
Vol 19 ◽  
pp. 346-361 ◽  
Author(s):  
Lloyd Hackel ◽  
Jon Rankin ◽  
Matt Walter ◽  
C Brent Dane ◽  
William Neuman ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Corrosion Cracking ◽  
Dry Storage

scholarly journals Detecting and imaging stress corrosion cracking in stainless steel, with application to inspecting storage canisters for spent nuclear fuel

2020 ◽  
Vol 109 ◽  
pp. 102180 ◽  
Author(s):  
Marcel C. Remillieux ◽  
Djamel Kaoumi ◽  
Yoshikazu Ohara ◽  
Marcie A. Stuber Geesey ◽  
Li Xi ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Corrosion Cracking

Crack Growth Rate Testing and Large Plate Demonstration Under Chloride-Induced Stress Corrosion Cracking Conditions in Stainless Steel Canisters for Storage of Spent Nuclear Fuel

2019 ◽  
Author(s):  
Poh-Sang Lam ◽  
Andrew J. Duncan ◽  
Lisa N. Ward ◽  
Robert L. Sindelar ◽  
Yun-Jae Kim ◽  
...  
Keyword(s):  
Growth Rate ◽  
Stainless Steel ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Compact Tension ◽  
Corrosion Cracking

Abstract Stress corrosion cracking may occur when chloride-bearing salts deposit and deliquesce on the external surface of stainless steel spent nuclear fuel storage canisters at weld regions with high residual stresses. Although it has not yet been observed, this phenomenon leads to a confinement concern for these canisters due to its potential for radioactive materials breaching through the containment system boundary provided by the canister wall during extended storage. The tests for crack growth rate have been conducted on bolt-load compact tension specimens in a setup designed to allow initially dried salt deposits to deliquesce and infuse to the crack front under conditions relevant to the canister storage environments (e.g., temperature and humidity). The test and characterization protocols are performed to provide bounding conditions in which cracking will occur. The results after 2- and 6-month exposure are examined in relation to previous studies in condensed brine and compared with other experimental data in the open literature. The knowledge gained from bolt-load compact tension testing is being applied to a large plate cut from a mockup commercial spent nuclear fuel canister to demonstrate the crack growth behavior induced from starter cracks machined in regions where the welding residual stress is expected. All these tests are conducted to support the technical basis for ASME Boiler and Pressure Vessel Section XI Code Case N-860.

scholarly journals Development of a numerical model for simulating stress corrosion cracking in spent nuclear fuel canisters

2021 ◽  
Vol 5 (1) ◽  
Author(s):  
Xin Wu ◽  
Fengwen Mu
Keyword(s):  
Numerical Model ◽  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Numerical Models ◽  
Corrosion Cracking ◽  
304 Stainless Steel ◽  
Research Progress ◽  
Nuclear Fuel Canisters

AbstractPrediction and detection of the chloride-induced stress corrosion cracking (CISCC) in Type 304 stainless steel spent nuclear fuel canisters are vital for the lifetime extension of dry storage canisters. This paper conducts a critical review that focuses on the numerical modeling and simulation on the research progress of the CISCC. The numerical models emphasizing the residual stress, susceptible microstructure, and corrosive environment are summarized individually. Meanwhile, the simulation studies on the role of hydrogen-assisted cracking are reviewed. Finally, a multi-physical numerical model, which combines the different fields is proposed based on our recent investigation.

An Approach to Model Abstraction of Stress Corrosion Cracking Damage in Management of Spent Nuclear Fuel and High-Level Waste

2013 ◽  
Author(s):  
Tae M. Ahn
Keyword(s):  
Crack Propagation ◽  
Stress Corrosion ◽  
Nuclear Fuel ◽  
Stress Corrosion Cracking ◽  
Spent Nuclear Fuel ◽  
Corrosion Cracking ◽  
Propagation Model ◽  
High Level Waste ◽  
Multiple Cracks ◽  
High Level

This paper presents an approach to assess stress corrosion cracking (SCC) damage of a canister for use in confinement management (extended dry storage or geological disposal) of radionuclides from spent nuclear fuel and high-level (radioactive) waste. Localized corrosion, mainly in pitting form and fabrication flaws, were analyzed as a possible precursor to SCC using field/laboratory data. This paper assesses single crack propagation over long time periods and estimates the potential maximum opening area resulting from multiple cracks. This crack propagation model was developed by the Sandia National Laboratories (SNL) for disposal under seismic conditions, and it appears to be conservative with respect to radionuclide releases through the opening area. The SNL model could be applied to the weld and various metals for both management applications. The conservative SNL approach could be used to estimate consequences of radionuclides dispersals, if a canister failed as the confinement barrier.

Sign in / Sign up

Export Citation Format

Share Document