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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Study of Mechanical Properties, Microstructure, and Residual Stresses of AISI 304/304L Stainless Steel Submerged Arc Weld for Spent Fuel Dry Storage Systems

The confinement boundaries of spent nuclear fuel (SNF) canisters are typically fusion welded. Welded microstructures, strain hardening, and residual stresses combined with a chemically aggressive, chloride-rich environment led to concerns that the welded canister may be susceptible to chloride-induced stress corrosion cracking (CISCC). A comprehensive understanding of the modification of stainless steel (SS) metallurgical and mechanical properties by fusion welding could accelerate the predictive analysis of CISCC susceptibility. This paper describes a submerged arc welding (SAW) procedure that was developed and qualified on 12.7 mm (0.5 in.) thick AISI 304/304L SS to produce joints in a way similar to actual SNF canister manufacturing. This procedure has the potential to reduce the production cost and weld CISCC susceptibility by using fewer welding passes and lower heat input than current industrial applications. Global and local mechanical behaviors and properties, as well as residual stress distributions on the welded joint, were studied. The results indicate that hardness values in the fusion zone (FZ) and heat-affected zone (HAZ) are slightly higher than that of the base metal. Strain localization was presented in the HAZ before the tensile stress reached its maximum value, and then it shifted to the FZ. The specimen finally broke in the FZ. High tensile residual stresses exhibited in the FZ and the nearby HAZ suggest the highest CISCC-susceptible spots. The maximum tensile residual stresses were along the welding direction, indicating that if cracks occur, they would be perpendicular to the welding direction. This study involved developing and qualifying a SAW procedure for SNF canister production. The new procedure yielded cost savings (SAW working efficiency increased by about 80%), improved mechanical properties, and presented moderate residual stresses. Analysis revealed that the welded joint’s low-stress and high-stress damage assessments may be affected by shifts in the strain localization spot under loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Lifetime estimation of a BWR core shroud in terms of IGSCC

The continued operation of aging Boiling Water Reactors (BWRs) worldwide requires gradually increasing cost of inspection, maintenance, and repair. Intergranular Stress Corrosion Cracking (IGSCC) in sensitized austenitic stainless steel piping first became a major issue for BWRs in the 1980s, resulting in recognition of the susceptibility of reactor internals to IGSCC. Shroud cracking identified in 1993–1994 confirmed that IGSCC of internals is a significant issue for BWRs. IGSCC is a time-dependent, material degradation process, which is caused and accelerated by the presence of residual stresses, material sensitization, irradiation, cold work, elevated temperature, and corrosive environments. This paper emphasizes the importance of accounting for corrosive environments, or more exactly, electrochemical phenomena in modelling IGSCC and predicting the service life of BWR in-vessel components, and stresses the necessity of performing such modeling not just for a single state point under full power conditions, but for the whole operating history of the reactor, including startups and shutdowns. Overall, this paper demonstrates that ignoring electrochemical considerations may result in underestimating component lifetime and lead to unnecessary expenses for inspection and repair.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2006

Slow Crack Growth and Fracture Toughness of Sapphire for the International Space Station Fluids and Combustion Facility

The fracture toughness, inert flexural strength, and slow crack growth parameters of the r- and a-planes of sapphire grown by the Heat Exchange Method were measured to qualify sapphire for structural use in the International Space Station. The fracture toughness in dry nitrogen, K(sub Ipb), was 2.31 +/- 0.12 MPa(square root of)m and 2.47 +/- 0.15 MPa(squre root of)m for the a- and r-planes, respectively. Fracture toughness measured in water via the operational procedure in ASTM C1421 was significantly lower, K(sub Ivb) = 1.95+/- 0.03 MPa(square root of)m, 1.94 +/- 0.07 and 1.77 +/- 0.13 MPa(square root of)m for the a- , m- and r-planes, respectively. The mean inert flexural strength in dry nitrogen was 1085 +/- 127 MPa for the r-plane and 1255 +/- 547 MPa for the a-plane. The power law slow crack growth exponent for testing in water was n = 21 +/- 4 for the r-plane and n (greater than or equal to) 31 for the a-plane. The power law slow crack growth coefficient was A = 2.81 x 10(exp -14) m/s x (MPa(squre root of)m)/n for the r-plane and A (approx. equals)2.06 x 10(exp -15) m/s x (MPa(square root of)m)/n for the a-plane. The r- and a-planes of sapphire are relatively susceptible to stress corrosion induced slow crack growth in water. However, failure occurs by competing modes of slow crack growth at long failure times and twinning for short failure time and inert environments. Slow crack growth testing needs to be performed at low failure stress levels and long failure times so that twinning does not affect the results. Some difficulty was encountered in measuring the slow crack growth parameters for the a-plane due to a short finish (i.e., insufficient material removal for elimination of the damage generated in the early grinding stages). A consistent preparation method that increases the Weibull modulus of sapphire test specimens and components is needed. This would impart higher component reliability, even if higher Weibull modulus is gained at the sacrifice of absolute strength of the component. The current specification frequently used for the preparation of sapphire test specimens and components (e.g., a "60/40" scratch-dig finish) is inadequate to avoid a short finish.

Salem, Jonathan A.↗
DOE OSTI2025

FLAW STABILITY ANALYSIS OF SURFACE CRACKS IN DOE STANDARD CANISTERS UNDER OPERATION LOADS AND WELDING RESIDUAL STRESSES

There are over 3000 commercial spent nuclear fuel (SNF) storage canisters that are made of stainless steel and planned for multi-purpose functions of SNF handling: interim storage, transportation, and ultimate disposal at a future SNF disposition location. To date, many multi-purpose canisters (MPC) are located in coastal regions for long-term storage. The canisters are fabricated by welding, and post-welding heat treatment is not required for relieving welding residual stresses (WRS). As a result, these canisters may be susceptible to chloride induced stress corrosion cracking (CISCC) due to the chloride-bearing marine salts. The flaw stability analysis of MPC canisters was performed at SRNL (i.e., PVP2016-4935 and PVP2016-63887) for through-wall flaws and for surface flaws respectively, where WRS was considered, and the failure assessment diagram (FAD)-based fracture mechanics method codified by API 579-1/ASME FFS-1-2007 Edition was adopted.

DOE standard canister↗
DOE OSTI2021

Preliminary Modeling of Chloride Deposition on Spent Nuclear Fuel Canisters in Dry Storage Relevant to Stress Corrosion Cracking

This study presents a method that can be used to gain information relevant to determining the corrosion risk for spent nuclear fuel (SNF) canisters during extended dry storage. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. This study uses well-developed computational fluid dynamics and particle-tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. The deposition rate is determined for a vertical canister system and a horizontal canister system, at various decay heat rates with a uniform particle size distribution, ranging from 0.25 to 25 µm, used as an input. In all cases, most of the dust entering the overpack passed through without depositing. Most of what was retained in the overpack was deposited on overpack surfaces (e.g., inlet and outlet vents); only a small fraction was deposited on the canister itself. These results are provided for generalized canister systems with a generalized input; as such, this paper is intended to demonstrate the technique. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Nuclear Energy Office of Spent Fuel Waste Science and Technology, which is tasked with doing research relevant to developing a sound technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

FY2022 Status Report: Cold Spray for Canister SCC Mitigation and Repair

This progress report describes work performed during FY22 at Sandia National Laboratories (SNL) to assess the corrosion performance of cold spray coatings to enable optimization of cold spray for the purposes of mitigation and/or repair of potentially susceptible regions, corrosion, or stress corrosion cracking (SCC) in austenitic stainless steel for spent nuclear fuel (SNF) storage. Of particular concern is SCC, by which a through-wall crack could potentially form in a canister outer wall over time intervals that may be shorter than possible dry storage times. In FY21, initial corrosion explorations of cold spray coating were evaluated and in FY22, an expanded set of cold spray coatings with in-depth analysis of post-exposure accelerated testing was explored. Additionally, relevant atmospheric exposure testing was carried out and initial results are presented herein. The corrosion attack from the accelerated testing and more realistic atmospheric exposures environments were compared to identify potentially deleterious factors for corrosion as well as help to understand the applicability of accelerated testing for cold spray optimization. This initial analysis will help to enable optimization of the corrosion resistance cold spray, one of the more promising coating and repair techniques, for potential application in an SNF environment. Learnings from both are summarized, and implications and future work are presented in this report.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Quasi-Extended Range Optical Frequency Domain Reflectometry for Product Pipelines

Throughout the United States, large natural gas and product transmission pipelines exist. These pipelines are susceptible to failure through corrosion and cracking due to internal and external factors. As these transmission pipeline age, the likelihood of a catastrophic failure increases. The ability to the monitor corrosion and wall thickness of these pipelines is paramount to reduce risk of disastrous failure and increase reliability as well as safety. Optical frequency domain reflectometry can monitor distributed temperature and strain measurements along the natural gas and product pipelines. This measurement technique can provide valuable information of pipe structural health through hoop strain changes due to pipe wall thinning, temperature changes due to gas leaks based on Joule-Thomson effect. Since these pipelines are typically buried, the depth of the pipeline results in a potential loss in the sensing range where the fiber is not physically monitoring the hoop strain of the pipeline. A different configuration of the interferometer will allow for the distributed fiber measurement to start atop the pipeline itself. This configuration results in no loss of sensing range and maintains the sensing resolution of a typically interferometer for an optical frequency domain reflectometry measurement. The quasi-extended range distributed hoop strain measurements were demonstrated on an active transmission pipeline.

Brister, Matthew↗
DOE OSTI2023

Irradiation-assisted cracking of SA508-304L weldments with 308L groove filler and 309L butter in hot water immersion constant rate extension tests

Low alloy SA508 bainitic steel is often welded to 300 series austenitic stainless steel in nuclear power plants. These dissimilar metal weldments use 309L butter applied to the low alloy steel and 308L groove filler material to join the 309L to austenite. High alloy steels (such as the 308L and 309L material used here) have been identified as having greater susceptibility to irradiation-assisted stress corrosion cracking (IASCC). Evidence is presented here of cracking under tensile loads in 309L weldment butter in LWR hydrothermal conditions that correlates irradiation induced changes at δ ferrite–γ austenite phase boundaries and γ–γ grain boundaries; specifically intergranular Cr 23 C 6 precipitates. Cracking is not observed in irradiated 308L groove filler under the same conditions. Both the 309L butter and 308L filler material have a duplex or mixed δ–γ phase microstructure due to solidification during the welding procedure. We have performed constant extension rate autoclave immersion tests in BWR normal water chemistry hot water (288 °C, 10 MPa, 2000 wt. ppb dissolved oxygen, <100 nS/cm conductivity, neutral pH) of proton irradiated (2 MeV, approximately 3 displacements per atom at a depth of 10 μm) tensile specimens. Proton irradiation to the dpa levels used here induced Cr 23 C 6 carbide precipitation at γ–γ grain boundaries and δ–γ phase boundaries in the 309L butter but not the 308L filler. Cracking was observed in the proton-irradiated volume of 309L at these interfaces in the vicinity of the Cr 23 C 6 precipitates. We hypothesis the SA508 is the source of carbon in the 309L butter. Cr depletion at γ–γ grain boundaries was also observed and may also be a root cause of cracking. Furthermore, the 309L butter is under significant tensile residual stress in the weldment that adds to applied tensile loads and this too may be a contributing factor.

309L weld butter↗
NASA NTRS1972

The effect of alloy composition on the mechanism of stress corrosion cracking of titanium alloys in aqueous environments

The effects of alloy composition on the aqueous stress corrosion of titanium alloys were studied with emphasis on determining the interrelations among composition, phase structure, and deformation and fracture properties of the alpha phase in alpha-beta alloys. Accomplishments summarized include the effects of alloy composition on susceptibility, and metallurgical mechanisms of stress-corrosion cracking.

Boyd, J. D.↗
NASA NTRS1994

Modeling time-dependent corrosion fatigue crack propagation in 7000 series aluminum alloys

Stress corrosion cracking and corrosion fatigue experiments were conducted with the susceptible S-L orientation of AA7075-T651, immersed in acidified and inhibited NaCl solution, to provide a basis for incorporating environmental effects into fatigue crack propagation life prediction codes such as NASA FLAGRO. This environment enhances da/dN by five to ten-fold compared to fatigue in moist air. Time-based crack growth rates from quasi-static load experiments are an order of magnitude too small for accurate linear superposition prediction of da/dN for loading frequencies above 0.001 Hz. Alternate methods of establishing da/dt, based on rising-load or ripple-load-enhanced crack tip strain rate, do not increase da/dt and do not improve linear superposition. Corrosion fatigue is characterized by two regimes of frequency dependence; da/dN is proportional to f(exp -1) below 0.001 Hz and to F(exp 0) to F(exp -0.1) for higher frequencies. Da/dN increases mildly both with increasing hold-time at K(sub max) and with increasing rise-time for a range of loading waveforms. The mild time-dependence is due to cycle-time-dependent corrosion fatigue growth. This behavior is identical for S-L nd L-T crack orientations. The frequency response of environmental fatigue in several 7000 series alloys is variable and depends on undefined compositional or microstructural variables. Speculative explanations are based on the effect of Mg on occluded crack chemistry and embritting hydrogen uptake, or on variable hydrogen diffusion in the crack tip process zone. Cracking in the 7075/NaCl system is adequately described for life prediction by linear superposition for prolonged load-cycle periods, and by a time-dependent upper bound relationship between da/dN and delta K for moderate loading times.

Mason, Mark E.↗
NASA NTRS2004

Electrochemical Impedance Spectroscopy of Alloys in a Simulated Space Shuttle Launch Environment

Type 304L stainless steel (304L SS) tubing is currently used in various supply lines that service the Orbiter at NASA's John F. Kennedy Space Center Launch Pads in Florida (USA). The atmosphere at the Space Shuffle launch site is very corrosive due to a combination of factors, such as the proximity of the Atlantic Ocean and the concentrated hydrochloric acid produced by the fuel combustion reaction in the solid rocket boosters. The acidic chloride environment is aggressive to most metals and causes severe pitting in many of the common stainless steel alloys such as 304L SS. Stainless steel tubing is susceptible to pitting corrosion that can cause cracking and rupture of both high-pressure gas and fluid systems. Outages in the systems where failures occur can impact the normal operation of the shuttle and launch schedules. The use of a more corrosion resistant tubing alloy for launch pad applications would greatly reduce the probability of failure, improve safety, lessen maintenance costs, and reduce downtime. A study which included ten alloys was undertaken to find a more corrosion resistant material to replace the existing 304L SS tubing. The study included atmospheric exposure at NASA's John F. Kennedy Space Center outdoor corrosion test site near the launch pads and electrochemical measurements in the laboratory which included DC techniques and electrochemical impedance spectroscopy (EIS). This paper presents the results from EIS measurements on three of the alloys: AL6XN (UN N08367), 254SMO (UNS S32l54), and 304L SS (UNS S30403). Type 304L SS was included in the study as a control. The alloys were tested in three electrolyte solutions which consisted of neutral 3.55% NaC1, 3.55% NaCl in O.1N HC1, and 3.55% NaCl in 1.ON HC1. The solutions were chosen to simulate environments that were expected to be less, similar, and more aggressive, respectively, than those present at the Space Shuttle launch pads. The results from the EIS measurements were analyzed to evaluate the corrosion susceptibility of the alloys and to predict the long-term corrosion performance of the subject materials. The results from the EIS measurements for the three alloys indicated that the higher-alloyed 254SMO and AL6XN exhibited a significantly improved resistance to corrosion than the 304L SS as the concentration of hydrochloric acid in the 3.55% NaC1 solution was increased. The polarization resistance values obtained from the EIS measurements were consistent with those from linear polarization measurements, and were indicative of the actual long-term corrosion performance of the alloys during a two-year atmospheric exposure study.

Calle, L. M.↗
DOE OSTI2021

Radiation Effects Related to Repaired BWR Core Shrouds

The material properties of the core shroud and supporting tie rod components are important factors for Boiling Water Reactor (BWR) life extension considerations. Given the variability of nickel and boron content found in 304 stainless steel shroud plates, and XM-19 and Inconel X-750 tie rod materials, this paper combines the latest spectral information in the vicinity of the core shroud to calculate hydrogen and helium gas production limits as well as atomic displacement damage spanning 80 full-power years (FPY) using the SPECTER code. Results indicate that, whereas boron has little effect on the atomic displacement damage, helium production from boron can be important and must be included in weldability assessments. Since the boron concentration is uncertain, helium generation is calculated assuming a composition range based on specifications and reported measurements. The results show that helium production caused by transmutations in the 304 SS shroud material are dominated by the concentration of boron and will have a significant impact on weldability after only a few years of service. For tie rods, the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) has been assessed by determining helium production in Inconel X-750 IASCC specimens irradiated in the Advanced Test Reactor (ATR) reactor and correlating with the helium produced in a BWR. As a tie rod material in a high thermal neutron flux environment, Inconel X-750 is more likely to be IASCC sensitive during service in a BWR due to helium than the XM-19 stainless steel alloy because of a greater nickel content and boron concentration. Weldability of the core shroud plate material, IASCC of the tie rod materials, and the gamma photon contribution to radiation damage are discussed.

radiation damage in materials, BWR↗
DOE OSTI2020

Investigation of Cold Spray as a Dry Storage Canister Repair and Mitigation Tool

The lack of a permanent spent nuclear fuel repository means utilities are faced with extending licenses for dry cask storage systems (DCSS) canisters far beyond their original 20-year license period. DCSS canisters are made from austenitic stainless steels (SSs) and fabricated using fusion welds. High heat input and melting associated with fabrication welds causes material around the weld to be susceptible to chloride-induced stress corrosion cracking (CISCC). The purpose of the work reported here is to investigate the use of cold spray for repair and mitigation of CISCC in DCSS canisters to ensure integrity far beyond their original license period. This report presents an analysis of cold spray for canister life extension, a summary of related efforts, relevant technology down selections, work accomplished, and future efforts relative to use of cold spray for repair and mitigation of CISCC in DCSS canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

In-service Oxidation and IASCC in High Fluence Baffle-Former Bolts Retrieved from a Westinghouse PWR

Baffle-former bolts (BFBs) are subjected to significant mechanical stress and neutron irradiation from the reactor core during the plant operation. Over the long operation period, these conditions lead to potential degradation and reduced load-carrying capacity of the bolts, and life extension of existing pressurized water reactors (PWRs) would only cause more damage to the bolt material. To this end, the Light Water Reactor Sustainability (LWRS) Program Materials Research Pathway (MRP) successfully harvested two high fluence BFBs from a Westinghouse two-loop downflow type PWR in 2016. In the same year, the two BFBs were received at the Westinghouse Churchill facility for inspection and specimen fabrication. The fabrication was completed in 2017 with specimens shipped to Oak Ridge National Laboratory (ORNL) for further testing. The objective of this project is to provide information that is integral to evaluating end of life microstructure and properties as a benchmark of international models developed for predicting radiation-induced swelling, segregation, precipitation, mechanical property degradation, and with the latest findings, susceptibility to irradiation-assisted stress corrosion cracking (IASCC).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗