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Corrosion Evaluation of Aluminum Alloys in Deionized Water

Description: Spent nuclear fuels from foreign and domestic research and test reactors being returned to SRS are now stored with other nuclear materials in the L-basin at the Savannah River Site (SRS). Recent efforts have consolidated the fuel storage systems and L-basin has become the SRS site for water storage of spent nuclear fuels. Corrosion surveillance of coupons in this basin is being performed to provide assurance of safe storage of spent fuel. This paper describes the highlights of recent studies on these aluminum coupons after immersion for more than 7 years in L-basin. Selected coupons were metallurgically characterized to establish the existence of general corrosion and pitting. Pitting was observed on galvanically coupled samples and also on intentionally creviced coupons, thus demonstrating that localized concentration cells were formed during the exposure period. In these cases, the susceptibility to pitting was not attributed to aggressive basin water chemistry but to local condition s (crevices and galvanic coupling) that allowed the development of oxygen and/or metal ion concentration cells that produced locally aggressive waters. General corrosion was also observed on some of the coupons that had not been treated to enhance the protective oxide prior to exposure in the basin water. These observations demonstrate that, even when the basin water chemistry is rigorously controlled, localized aggressive conditions can develop. Although this demonstration does not suggest significant deterioration of the stored spent nuclear fuels, it does illustrate the potential for corrosion induced degradation and thus the importance of a routine surveillance program.
Date: September 24, 2004
Creator: Vormelker, Philip R. & Duncan, Andrew J.

Corrosion Evaluation of INTEC Waste Storage Tank WM-182

Description: Irradiated nuclear fuel has been stored and reprocessed at the Idaho National Engineering and Environmental Laboratory since 1953 using facilities located at the Idaho Nuclear Technology and Engineering Center (INTEC). This reprocessing produced radioactive liquid waste which was stored in the Tank Farm. The INTEC Tank Farm consists of eleven vaulted 300,000-gallon underground tanks including Tank WM-182. Tank WM-182 was put into service in 1955, has been filled four times, and has contained aluminum and zirconium fuel reprocessing wastes as well as sodium bearing waste. A program to monitor corrosion in the waste tanks was initiated in 1953 when the first of the eleven Tank Farm tanks was placed in service. Austenitic stainless steel coupons representative of the materials of construction of the tanks are used to monitor internal tank corrosion. This report documents the final inspection of the WM-182 corrosion coupons. Physical examination of the welded corrosion test coupons exposed to the tank bottom conditions of Tank WM-182 revealed very light uniform corrosion. Examination of the external surfaces of the extruded pipe samples showed very light uniform corrosion with slight indications of preferential attack parallel to extrusion marks and start of end grain attack of the cut edges. These indications were only evident when examined under stereo microscope at magnifications of 20X and above. There were no definite indications of localized corrosion, such as cracking, pitting, preferential weld attack, or weld heat affected zone attack on either the welded or extruded coupons. Visual examination of the coupon support cables, where they were not encased in plastic, failed to reveal any indication of liquid-liquid interface attack of any crevice corrosion. Based on the WM-182 coupon evaluations, which have occurred throughout the life of the tank, the metal loss from the tank wall due to uniform corrosion is not expected to ...
Date: November 1, 1999
Creator: Dirk, W. J. & Anderson, P. A.

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

Description: As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide ...
Date: September 1, 2012
Creator: Wertsching, A K


Description: Extrusions of aluminum alloy powder products were obtained from several sources and evaluated for corrosion resistance to high-temperature (260-- 350 deg C) water. Several types of tubing impact-extruded by ALCOA were tested. The stronger tabing (M655) failed very rapidly. The weaker tubing suffered extensive localized surface attack and penetration of the corrosion attack along the extrusion direction after prolonged ( approximates 3 months) exposure to 290 deg C water. A precorrosion heat treatment was effective in reducing both types of attack on the weaker tubing. Armour Research Foundation supplied several types of tubing extraded through a bridge die. All tubes failed on prolonged ( approximates 8 months) corrosion in 290 deg C water at the longitudinal bond lines. These lines were formed by the rejoining of the metal streams passing over the mandrel supports in the die during extrusion. Directly extruded tubing supplied by the Torrance Brass Company also failed on extended exposure to 290 deg C water. Many experimental rod extrusions (from Armour Research Foundation and Trefimetaux) exhibited corrosion resistance to static 290 deg C water equivalent to that of wrought alloys. The Trefimetaux specimens were also tested in rapidly flowing water at 315 deg C. Under these conditions a corrosion rate significantly faster than for the wrought alloy was measured. (auth)
Date: November 1, 1963
Creator: Draley, J.E.; Ruther, W.E. & Greenberg, S.

Corrosion experiments on stainless steels used in dry storage canisters of spent nuclear fuel

Description: Nonradioactive (cold) experiments have been set up in the Idaho Chemical Processing Plant (ICPP)-1634, and radioactive (hot) experiments have been set up in the Irradiated Fuel Storage Facility (IFSF) at ICPP. The objective of these experiments is to provide information on the interactions (corrosion) between the spent nuclear fuel currently stored at the ICPP and the dry storage canisters and containment materials in which this spent fuel will be stored for the next several decades. This information will be used to help select canister materials that will retain structural integrity over this period within economic, criticality, and other constraints. The two purposes for Dual Purpose Canisters (DPCs) are for interim storage of spent nuclear fuel and for shipment to a final geological repository. Information on how corrosion products, sediments, and degraded spent nuclear fuel may corrode DPCs will be required before the DPCs will be allowed to be shipped out of the State of Idaho. The information will also be required by the Nuclear Regulatory Commission (NRC) to support the licensing of DPCs. Stainless steels 304L and 316L are the most likely materials for dry interim storage canisters. Welded stainless steel coupons are used to represent the canisters in both hot and cold experiments.
Date: September 1, 1996
Creator: Ryskamp, J.M.; Adams, J.P.; Faw, E.M. & Anderson, P.A.

Corrosion fatigue crack growth in clad low-alloy steel. Part 2, Water flow rate effects in high sulfur plate steel

Description: Corrosion fatigue crack propagation tests were conducted on a high- sulfur ASTM A302-B plate steel overlaid with weld-deposited Alloy EN82H cladding. The specimens featured semi-elliptical surface cracks penetrating approximately 6.3 mm of cladding into the underlying steel. The initial crack sizes were relatively large with surface lengths of 22.8--27.3 mm, and depths of 10.5--14.1 mm. The experiments were initiated in a quasi-stagnant low-oxygen (O{sub 2} < 10 ppb) aqueous environment at 243{degrees}C, under loading conditions ({Delta}K, R, cyclic frequency) conducive to environmentally-assisted cracking (EAC) under quasi-stagnant conditions. Following fatigue testing under quasi-stagnant conditions where EAC was observed, the specimens were then fatigue tested under conditions where active water flow of either 1.7 m/sec. or 4.7 m/sec. was applied parallel to the crack. Earlier experiments on unclad surface-cracked specimens of the same steel exhibited EAC under quasi- stagnant conditions, but water flow rates at 1.7 m/sec. and 5.0 m/sec. parallel to the crack mitigated EAC. In the present experiments on clad specimens, water flow at approximately the same as the lower of these velocities did not mitigate EAC, and a free stream velocity approximately the same as the higher of these velocities resulted in sluggish mitigation of EAC. The lack of robust EAC mitigation was attributed to the greater crack surface roughness in the cladding interfering with flow induced within the crack cavity. An analysis employing the computational fluid dynamics code, FIDAP, confirmed that frictional forces associated with the cladding crack surface roughness reduced the interaction between the free stream and the crack cavity.
Date: April 1, 1996
Creator: James, L.A; Lee, H.B.; Wire, G.L.; Novak, S.R. & Cullen, W.H.

Corrosion fatigue crack growth in clad low-alloy steels: Part 1, medium-sulfur forging steel

Description: Corrosion fatigue crack propagation tests were conducted on a medium- sulfur ASTM A508-2 forging steel overlaid with weld-deposited Alloy EN82H cladding. The specimens featured semi-elliptical surface cracks penetrating approximately 6.3 mm of cladding into the underlying steel. The initial crack sizes were relatively large with surface lengths of 30.3--38.3 mm, and depths of 13.1--16.8 mm. The experiments were conducted in a quasi-stagnant low-oxygen (O{sub 2} < 10 ppb) aqueous environment at 243{degrees}C, under loading conditions ({Delta}K, R, and cyclic frequency) conductive to environmentally-assisted cracking (EAC) in higher-sulfur steels under quasi-stagnant conditions. Earlier experiments on unclad compact tension specimens of this heat of steel did not exhibit EAC, and the present experiments on semi-elliptical surface cracks penetrating cladding also did not exhibit EAC.
Date: April 1, 1996
Creator: James, L.A.; Poskie, T.J.; Auten, T.A & Cullen, W.H.

Corrosion fatigue of alloys 600 and 690 in simulated LWR environments

Description: Crack growth data were obtained on fracture-mechanics specimens of Alloys 600 and 690 to investigate environmentally assisted cracking (EAC) in simulated boiling water reactor and pressurized water reactor environments at 289 and 320 C. Preliminary information was obtained on the effect of temperature, load ratio, stress intensity (K), and the dissolved-oxygen and -hydrogen concentrations of the water on EAC. Specimens of Type 316NG and sensitized Type 304 stainless steel (SS) were included in several of the experiments to assess the behavior of these materials and Alloy 600 under the same water chemistry and loading conditions. The experimental data are compared with predictions from an Argonne National Laboratory (ANL) model for crack growth rates (CGRs) of SSs in water and the ASME Code Section 11 correlation for CGRs in air at the K{sub max} and load-ratio values in the various tests. The data for all of the materials were bounded by ANL model predictions and the ASME Section 11 ``air line.``
Date: April 1996
Creator: Ruther, W. E.; Soppett, W. K. & Kassner, T. F.

Corrosion in a temperature gradient

Description: High temperature corrosion limits the operation of equipment used in the Power Generation Industry. Some of the more destructive corrosive attack occurs on the surfaces of heat exchangers, boilers, and turbines where the alloys are subjected to large temperature gradients that cause a high heat flux through the accumulated ash, the corrosion product, and the alloy. Most current and past corrosion research has, however, been conducted under isothermal conditions. Research on the thermal-gradient-affected corrosion of various metals and alloys is currently being studied at the Albany Research Center’s SECERF (Severe Environment Corrosion and Erosion Research Facility) laboratory. The purpose of this research is to verify theoretical models of heat flux effects on corrosion and to quantify the differences between isothermal and thermal gradient corrosion effects. The effect of a temperature gradient and the resulting heat flux on corrosion of alloys with protective oxide scales is being examined by studying point defect diffusion and corrosion rates. Fick’s first law of diffusion was expanded, using irreversible thermodynamics, to include a heat flux term – a Soret effect. Oxide growth rates are being measured for the high temperature corrosion of cobalt at a metal surface temperature of 900ºC. Corrosion rates are also being determined for the high temperature corrosion of carbon steel boiler tubes in a simulated waste combustion environment consisting of O2, CO2, N2, and water vapor. Tests are being conducted both isothermally and in the presence of a temperature gradient to verify the effects of a heat flux and to compare to isothermal oxidation.
Date: January 1, 2003
Creator: Covino, Bernard S., Jr.; Holcomb, Gordon R.; Cramer, Stephen D.; Bullard, Sophie J.; Ziomek-Moroz, Margaret & White, M.L. (Convanta)

Corrosion in ICPP fuel storage basins

Description: The Idaho Chemical Processing Plant currently stores irradiated nuclear fuel in fuel storage basins. Historically, fuel has been stored for over 30 years. During the 1970`s, an algae problem occurred which required higher levels of chemical treatment of the basin water to maintain visibility for fuel storage operations. This treatment led to higher levels of chlorides than seen previously which cause increased corrosion of aluminum and carbon steel, but has had little effect on the stainless steel in the basin. Corrosion measurements of select aluminum fuel storage cans, aluminum fuel storage buckets, and operational support equipment have been completed. Aluminum has exhibited good general corrosion rates, but has shown accelerated preferential attack in the form of pitting. Hot dipped zinc coated carbon steel, which has been in the basin for approximately 40 years, has shown a general corrosion rate of 4 mpy, and there is evidence of large shallow pits on the surface. A welded Type 304 stainless steel corrosion coupon has shown no attack after 13 years exposure. Galvanic couples between carbon steel welded to Type 304 stainless steel occur in fuel storage yokes exposed to the basin water. These welded couples have shown galvanic attack as well as hot weld cracking and intergranular cracking. The intergranular stress corrosion cracking is attributed to crevices formed during fabrication which allowed chlorides to concentrate.
Date: September 1, 1993
Creator: Dirk, W.J.

Corrosion in Non-Hermetic Microelectronic Devices

Description: Many types of integrated and discrete microelectronic devices exist in the enduring stockpile. In the past, most of these devices have used conventional ceramic hermetic packaging (CHP) technology. Sometime in the future, plastic encapsulated microelectronic (PEM) devices will almost certainly enter the inventory. In the presence of moisture, several of the aluminum-containing metallization features common to both types of packaging become susceptible to atmospheric corrosion (Figure 1). A breach in hermeticity (e.g., due to a crack in the ceramic body or lid seal) could allow moisture and/or contamination to enter the interior of a CHP device. For PEM components, the epoxy encapsulant material is inherently permeable to moisture. A multi-year project is now underway at Sandia to develop the knowledge base and analytical tools needed to quantitatively predict the effect of corrosion on microelectronic performance and reliability. The issue of corrosion-induced failure surfaced twice during the past year because cracks were found in their ceramic bodies of two different CHP devices: the SA371 1/3712 MOSFET and the SA3935 ASIC (acronym for A Simple Integrated Circuit). Because of our inability to perform a model-based prediction at that time, the decision was made to determine the validity of the corrosion concern for these specific situations by characterizing the expected environment and assessing its relative degree of corrosivity. The results of this study are briefly described in this paper along with some of the advancements made with the predictive model development.
Date: March 16, 1999
Creator: Braithwaite, J. W. & Sorensen, N. R.

Corrosion in Supercritical carbon Dioxide: Materials, Environmental Purity, Surface Treatments, and Flow Issues

Description: The supercritical CO{sub 2} Brayton cycle is gaining importance for power conversion in the Generation IV fast reactor system because of its high conversion efficiencies. When used in conjunction with a sodium fast reactor, the supercritical CO{sub 2} cycle offers additional safety advantages by eliminating potential sodium-water interactions that may occur in a steam cycle. In power conversion systems for Generation IV fast reactors, supercritical CO{sub 2} temperatures could be in the range of 30°C to 650°C, depending on the specific component in the system. Materials corrosion primarily at high temperatures will be an important issue. Therefore, the corrosion performance limits for materials at various temperatures must be established. The proposed research will have four objectives centered on addressing corrosion issues in a high-temperature supercritical CO{sub 2} environment: Task 1: Evaluation of corrosion performance of candidate alloys in high-purity supercritical CO{sub 2}: The following alloys will be tested: Ferritic-martensitic Steels NF616 and HCM12A, austenitic alloys Incoloy 800H and 347 stainless steel, and two advanced concept alloys, AFA (alumina forming austenitic) steel and MA754. Supercritical CO{sub 2} testing will be performed at 450°C, 550°C, and 650°C at a pressure of 20 MPa, in a test facility that is already in place at the proposing university. High purity CO{sub 2} (99.9998%) will be used for these tests. Task 2: Investigation of the effects of CO, H{sub 2}O, and O{sub 2} impurities in supercritical CO{sub 2} on corrosion: Impurities that will inevitably present in the CO{sub 2} will play a critical role in dictating the extent of corrosion and corrosion mechanisms. These effects must be understood to identify the level of CO{sub 2} chemistry control needed to maintain sufficient levels of purity to manage corrosion. The individual effects of important impurities CO, H{sub 2}O, and O{sub 2} will be investigated by adding them ...
Date: December 10, 2013
Creator: Sridharan, Kumar & Anderson, Mark

Corrosion in supercritical fluids

Description: Integrated studies were carried out in the areas of corrosion, thermodynamic modeling, and electrochemistry under pressure and temperature conditions appropriate for potential applications of supercritical fluid (SCF) extractive metallurgy. Carbon dioxide and water were the primary fluids studied. Modifiers were used in some tests; these consisted of 1 wt% water and 10 wt% methanol for carbon dioxide and of sulfuric acid, sodium sulfate, ammonium sulfate, and ammonium nitrate at concentrations ranging from 0.00517 to 0.010 M for the aqueous fluids. The materials studied were Types 304 and 316 (UNS S30400 and S31600) stainless steel, iron, and AISI-SAE 1080 (UNS G10800) carbon steel. The thermodynamic modeling consisted of development of a personal computer-based program for generating Pourbaix diagrams at supercritical conditions in aqueous systems. As part of the model, a general method for extrapolating entropies and related thermodynamic properties from ambient to SCF conditions was developed. The experimental work was used as a tool to evaluate the predictions of the model for these systems. The model predicted a general loss of passivation in iron-based alloys at SCF conditions that was consistent with experimentally measured corrosion rates and open circuit potentials. For carbon-dioxide-based SCFs, measured corrosion rates were low, indicating that carbon steel would be suitable for use with unmodified carbon dioxide, while Type 304 stainless steel would be suitable for use with water or methanol as modifiers.
Date: May 1, 1996
Creator: Propp, W. A.; Carleson, T. E.; Wai, Chen M.; Taylor, P. R.; Daehling, K. W.; Huang, Shaoping et al.