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.

FeCrAl Alloys for Accident Tolerant Fuel Cladding in Light Water Reactors

2016 ◽  
Author(s):  
Raul B. Rebak ◽  
Kurt A. Terrani ◽  
Russ M. Fawcett
Keyword(s):  
Fuel Assembly ◽  
Department Of Energy ◽  
Normal Operation ◽  
Fuel System ◽  
Light Water Reactors ◽  
Light Water ◽  
Oak Ridge ◽  
Operation Conditions ◽  
Fecral Alloys ◽  
Water Reactors

The goal of the U.S. Department of Energy (DOE) Accident Tolerant Fuel Program (ATF) for light water reactors (LWR) is to identify alternative fuel system technologies to further enhance the safety of commercial nuclear power plants. An ATF fuel system would endure loss of cooling in the reactor for a considerably longer period of time than the current systems. The General Electric (GE) and Oak Ridge National Laboratory (ORNL) ATF design concept utilizes an iron-chromium-aluminum (FeCrAl) alloy material as fuel rod cladding in combination with uranium dioxide (UO2) fuel pellets currently in use, resulting in a fuel assembly that leverages the performance of existing/current LWR fuel assembly designs and infrastructure with improved accident tolerance. Significant testing was performed in the last three years to characterize FeCrAl alloys for cladding applications, both under normal operation conditions of the reactor and under accident conditions. This article is a state of the art description of the concept.

Used Fuel Management System Architecture and Interface Analyses

2013 ◽  
Author(s):  
Mark Nutt ◽  
Robert Howard ◽  
Ingrid Busch ◽  
Joe Carter ◽  
Alexcia Delley ◽  
...  
Keyword(s):  
Management System ◽  
Thermal Analyses ◽  
The United States ◽  
Department Of Energy ◽  
System Level ◽  
Fuel Management ◽  
Simulation Tools ◽  
Dry Storage ◽  
Water Reactors ◽  
Geologic Media

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications.

scholarly journals IRRADIATION EFFECTS ON ZR-2.5NB IN POWER REACTORS

2016 ◽  
Vol 5 (1) ◽  
pp. 17-36 ◽  
Author(s):  
Carol Song
Keyword(s):  
Mechanical Properties ◽  
Operating Experience ◽  
Zirconium Alloys ◽  
Light Water Reactors ◽  
Irradiation Effects ◽  
Light Water ◽  
Current State ◽  
Structural Applications ◽  
Water Reactors ◽  
Reactor Operating

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, “High Power Channel-type Reactor”) reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr-2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge.

Tritium Production and Permeation in High-Temperature Reactor Systems

2013 ◽  
Author(s):  
Piyush Sabharwall ◽  
Hans Schmutz ◽  
Carl Stoots ◽  
George Griffith
Keyword(s):  
Large Scale ◽  
Fusion Reaction ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Ternary Fission ◽  
Light Water Reactors ◽  
Tritium Production ◽  
High Temperature Reactor ◽  
Water Reactors ◽  
Tritium Permeation

Tritium (H13) is a radioactive isotope of hydrogen formed by ternary fission events (rare emissions of three nuclides rather than two during a fission) and neutron absorption (and subsequent decay) of predecessor radionuclides, particularly 6Li and 7Li. Also in fusion, the concept of breeding tritium during the fusion reaction is of significance for the future needs of a large-scale fusion power plant. Tritium is of special interest among the fission products created in next-generation nuclear reactors such as gas cooled reactors and molten salt reactors, because of the large quantities produced when compared with conventional light-water reactors (LWR) and the higher temperatures of operation for these systems enhances permeation. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for mitigation of permeation of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented along with mitigation strategies for tritium permeation.

Small Modular Molten Salt Reactor (SM-MSR)

2011 ◽  
Author(s):  
Piyush Sabharwall ◽  
Eung Soo Kim ◽  
Michael Mckellar ◽  
Mike Patterson
Keyword(s):  
Molten Salt ◽  
Thermal Power ◽  
The United States ◽  
Economic Benefits ◽  
Electricity Production ◽  
Molten Salt Reactor ◽  
Light Water Reactors ◽  
Reactor Outlet ◽  
Light Water ◽  
Water Reactors

The strategic goal of the Small Modular Molten Salt Reactor (SM-MSR) is to broaden the environmental and economic benefits of nuclear energy in the United States by producing power to meet growing energy demands and demonstrating its applicability to market sectors not being served by light water reactors. Of primary importance is demonstrating that the SM-MSR can be operated safely and compete economically with larger reactors. Reactor outlet temperatures (ROTs) of SM-MSRs will likely be much higher (around 700°C) than light water reactors, thereby increasing the efficiency of electricity production and potentially providing process heat for industrial applications, which will help offset the higher per kilowatt costs generally associated with smaller reactors, making the SM-MSR more economically competitive. This paper compares thermal power cycles for given ROT, compares thermal performance using figure of merits and sensitivity study and discusses the comparative advantages of SM-MSRs.

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