scholarly journals Acid dissolution behavior of ferritic FeCrAl tubes candidates for nuclear fuel cladding.

CORROSION ◽  
10.5006/3965 ◽  
2021 ◽  
Author(s):  
Raul Rebak ◽  
Liang Yin ◽  
Timothy Jurewicz ◽  
Andrew Hoffman
Keyword(s):  
Fuel Cycle ◽  
Austenitic Stainless Steels ◽  
Zirconium Alloys ◽  
Dissolution Behavior ◽  
Acid Dissolution ◽  
Alumina Layer ◽  
Electrochemical Tests ◽  
High Temperature Reaction ◽  
Mineral Acids ◽  
Fecral Alloys

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum -FeCrAl- alloys, which offer extraordinary resistance to high temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve the dissolution of the fuel rods in mineral acids. Little or nothing is known on the dissolution of FeCrAl alloys in common mineral acids, therefore the objective of this research was to study the dissolution of typical cladding tubing having two compositions of FeCrAl (APMT & C26M) in three acids (H2SO4, HNO3 & HCl) as a function of the temperature using both standard ASTM immersion tests as well as electrochemical tests. The dissolution behavior of the FeCrAl alloys is compared to the dissolution capability of other traditional nuclear materials such as austenitic stainless steels (304SS & 316SS) and austenitic nickel alloys (Alloy 600 and Hastelloy C-276). Results show that both C26M and APMT have a higher dissolution capability in the studied mineral acids, which will be beneficial for reprocessing procedures.

scholarly journals Resistance of Ferritic FeCrAl Alloys to Stress Corrosion Cracking for Light Water Reactor Fuel Cladding Applications

CORROSION ◽  
10.5006/3632 ◽  
2020 ◽  
Vol 76 (11) ◽  
Author(s):  
Raul B. Rebak ◽  
Liang Yin ◽  
Peter L. Andresen
Keyword(s):  
High Temperature ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Fuel Cycle ◽  
Austenitic Stainless Steels ◽  
Corrosion Cracking ◽  
Fuel Cladding ◽  
Light Water ◽  
Water Reactor ◽  
Fecral Alloys

Since 2011, the international nuclear materials community has been engaged in finding replacements for zirconium alloys fuel cladding for light water reactors. Iron-chromium-aluminum (FeCrAl) alloys are cladding candidates because they have high strength at high temperature and an extraordinary resistance to attack by superheated steam in the event of a loss of coolant accident. As FeCrAl alloys have never been used in nuclear reactors, it is important to characterize their behavior in the entire fuel cycle. Stress corrosion cracking (SCC) studies were conducted for two FeCrAl alloys (APMT and C26M) in typical simulated boiling water reactor conditions at 288°C containing either dissolved hydrogen or oxygen. Crack propagation studies showed that both ferritic FeCrAl alloys were resistant to SCC at stress intensities below 40 MPa√m. Current results for FeCrAl confirm previous findings for Fe-Cr alloys showing that ferritic stainless alloys are generally much more resistant to high-temperature water SCC than austenitic stainless steels.

Hydrogen Diffusion in FeCrAl Alloys for Light Water Reactors Cladding Applications

2016 ◽  
Author(s):  
Raul B. Rebak ◽  
Young-Jin Kim
Keyword(s):  
Hydrogen Diffusion ◽  
Zirconium Alloys ◽  
The United States ◽  
Department Of Energy ◽  
Light Water Reactors ◽  
Alumina Layer ◽  
Fecral Alloys ◽  
Water Reactors ◽  
Tritium Permeation ◽  
Storage Pools

There is a worldwide effort to develop nuclear fuels that are resistant to accidents such as loss of coolant in the reactor and the storage pools. In the United States, the Department of Energy is teaming with fuel vendors to develop accident tolerant fuels (ATF), which will resist the lack of cooling for longer periods of times than the current zirconium alloy - uranium dioxide system. General Electric (GE) and its partners is proposing to replace zirconium alloys cladding with an Iron-Chromium-Aluminum (FeCrAl) alloy such as APMT, since they are highly resistant to attack by steam up to the melting point of the alloy. FeCrAl alloys do not react with hydrogen to form stable hydrides as zirconium alloys do. Therefore, it is possible that more tritium may be released to the coolant with the use of FeCrAl cladding. This work discusses the formation of an alumina layer on the surface of APMT cladding as an effective barrier for tritium permeation from the fuel to the coolant across the cladding wall.

scholarly journals Effect of Mn Content and Solution Annealing Temperature on the Corrosion Resistance of Stainless Steel Alloys

2014 ◽  
Vol 2014 ◽  
pp. 1-8 ◽  
Author(s):  
Ihsan-ul-Haq Toor
Keyword(s):  
Corrosion Resistance ◽  
Annealing Temperature ◽  
Electrochemical Impedance ◽  
Austenitic Stainless Steels ◽  
Electrochemical Tests ◽  
Linear Polarization Resistance ◽  
Ni Content ◽  
Different Temperatures ◽  
Acidic Chloride ◽  
Mn Content

The corrosion behavior of two specially designed austenitic stainless steels (SSs) having different Nickel (Ni) and Manganese (Mn) contents was investigated. Prior to electrochemical tests, SS alloys were solution-annealed at two different temperatures, that is, at 1030°C for 2 h and 1050°C for 0.5 h. Potentiodynamic polarization (PD) tests were carried out in chloride and acidic chloride, whereas linear polarization resistance (LPR) and electrochemical impedance spectroscopy (EIS) was performed in 0.5 M NaCl solution at room temperature. SEM/EDS investigations were carried out to study the microstructure and types of inclusions present in these alloys. Experimental results suggested that the alloy with highest Ni content and annealed at 1050°C/0.5 hr has the highest corrosion resistance.

General Corrosion of Chromium-Coated Zirconium- and Titanium-Based Alloys in Supercritical Water at 500 °C

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
K. Khumsa-Ang ◽  
M. Edwards ◽  
S. Rousseau
Keyword(s):  
Supercritical Water ◽  
Fuel Cycle ◽  
Neutron Cross Section ◽  
Zirconium Alloys ◽  
Oxide Thickness ◽  
General Corrosion ◽  
Coating Materials ◽  
Grade 5 ◽  
Oxide Phases ◽  
Enriched Uranium

Abstract The 300 MWel small Canadian supercritical water-cooled reactor (SCWR), which is a scaled-down version of the original 1200 MWel concept, has a smaller core, uses low enriched uranium fuel instead of a plutonium–thorium fuel, and features a lower (maximum) cladding temperature of 500 °C. The lower cladding temperature may permit the use of different alloys, including zirconium alloys, which had been ruled out as candidates for the Canadian SCWR, whose cladding temperature may reach 850 °C. The potential to use zirconium alloys is exciting because they have a low neutron cross section, which in turn means that fewer neutrons are lost, and the fuel can be used more efficiently. One advantage, for example,, is that the fuel cycle can be lengthened. In this paper, we report on the results of corrosion experiments used to screen zirconium- and titanium-based alloys as well as corrosion-resistant coating materials such as Cr and Al as potential candidates for fuel cladding in the small Canadian SCWR. These experiments were conducted in a refreshed autoclave in deaerated supercritical water at 500 °C and 23.5 MPa. After exposure, the weight gain was measured, and the oxide thickness and the oxide phases were examined. Of all materials, the coated and uncoated Ti-grade 2 and Ti-grade 5 alloys met our screening qualification criteria, however, Al/Cr-coated zirconium coupons showed notable improvement and will be explored further in future testing.

scholarly journals Understanding the nature of atmospheric acid processing of mineral dusts in supplying bioavailable phosphorus to the oceans

2016 ◽  
Vol 113 (51) ◽  
pp. 14639-14644 ◽  
Author(s):  
Anthony Stockdale ◽  
Michael D. Krom ◽  
Robert J. G. Mortimer ◽  
Liane G. Benning ◽  
Kenneth S. Carslaw ◽  
...  
Keyword(s):  
Dust Particles ◽  
Dissolution Behavior ◽  
Acid Dissolution ◽  
Airborne Dust ◽  
Bioavailable Phosphorus ◽  
Surface Ocean ◽  
Acid Processing ◽  
Water Ph ◽  
Mineral Dusts ◽  
Direct Proportionality

Acidification of airborne dust particles can dramatically increase the amount of bioavailable phosphorus (P) deposited on the surface ocean. Experiments were conducted to simulate atmospheric processes and determine the dissolution behavior of P compounds in dust and dust precursor soils. Acid dissolution occurs rapidly (seconds to minutes) and is controlled by the amount of H+ions present. For H+< 10−4mol/g of dust, 1–10% of the total P is dissolved, largely as a result of dissolution of surface-bound forms. At H+> 10−4mol/g of dust, the amount of P (and calcium) released has a direct proportionality to the amount of H+consumed until all inorganic P minerals are exhausted and the final pH remains acidic. Once dissolved, P will stay in solution due to slow precipitation kinetics. Dissolution of apatite-P (Ap-P), the major mineral phase in dust (79–96%), occurs whether calcium carbonate (calcite) is present or not, although the increase in dissolved P is greater if calcite is absent or if the particles are externally mixed. The system was modeled adequately as a simple mixture of Ap-P and calcite. P dissolves readily by acid processes in the atmosphere in contrast to iron, which dissolves more slowly and is subject to reprecipitation at cloud water pH. We show that acidification can increase bioavailable P deposition over large areas of the globe, and may explain much of the previously observed patterns of variability in leachable P in oceanic areas where primary productivity is limited by this nutrient (e.g., Mediterranean).

scholarly journals Hydrogen Isotopes Permeation in Clean or Unoxidized FeCrAl Alloys: A Review

2022 ◽  
Author(s):  
Yogendra S. Garud ◽  
Andrew K. Hoffman ◽  
Raul B. Rebak
Keyword(s):  
Hydrogen Permeation ◽  
Austenitic Stainless Steels ◽  
Major Elements ◽  
Nuclear Industry ◽  
Department Of Energy ◽  
Loss Of Coolant Accident ◽  
Austenitic Alloys ◽  
Fecral Alloys ◽  
Water Reactors ◽  
Accident Scenarios

AbstractThe US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.

Radiation Damage Studies with the HVEM

1972 ◽  
Vol 30 ◽  
pp. 680-681
Author(s):  
J. J. Laidler ◽  
B. Mastel
Keyword(s):  
Radiation Damage ◽  
Stainless Steels ◽  
Volume Expansion ◽  
Austenitic Stainless Steels ◽  
Test Facility ◽  
Fast Breeder Reactors ◽  
Breeder Reactors ◽  
Under Construction ◽  
Development Laboratory ◽  
Insight Into

One of the major materials problems encountered in the development of fast breeder reactors for commercial power generation is the phenomenon of swelling in core structural components and fuel cladding. This volume expansion, which is due to the retention of lattice vacancies by agglomeration into large polyhedral clusters (voids), may amount to ten percent or greater at goal fluences in some austenitic stainless steels. From a design standpoint, this is an undesirable situation, and it is necessary to obtain experimental confirmation that such excessive volume expansion will not occur in materials selected for core applications in the Fast Flux Test Facility, the prototypic LMFBR now under construction at the Hanford Engineering Development Laboratory (HEDL). The HEDL JEM-1000 1 MeV electron microscope is being used to provide an insight into trends of radiation damage accumulation in stainless steels, since it is possible to produce atom displacements at an accelerated rate with 1 MeV electrons, while the specimen is under continuous observation.

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