scholarly journals Improving Nuclear Power Plant Safety with FeCrAl Alloy Fuel Cladding

MRS Advances ◽  
2017 ◽  
Vol 2 (21-22) ◽  
pp. 1217-1224 ◽  
Author(s):  
Raul B. Rebak ◽  
Kurt A. Terrani ◽  
William P. Gassmann ◽  
John B. Williams ◽  
Kevin L. Ledford
Keyword(s):  
Nuclear Power ◽  
National Laboratory ◽  
Alloy System ◽  
Department Of Energy ◽  
Normal Operation ◽  
Oak Ridge ◽  
Plant Safety ◽  
Operation Conditions ◽  
Fecral Alloy ◽  
Nuclear Power Plant Safety

ABSTRACTThe US Department of Energy (DOE) is partnering with fuel vendors to develop enhanced accident tolerant nuclear fuels for Generation III water cooled reactors. In comparison with the standard current uranium dioxide and zirconium alloy system UO2-Zr), the proposed alternative accident tolerant fuel (ATF) should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General Electric, Oak Ridge National Laboratory and their partners have proposed to replace zirconium based alloy cladding in current commercial power reactors with an iron-chromium-aluminum (FeCrAl) alloy cladding such as APMT. The use of FeCrAl alloys will greatly reduce the risk of operating the power reactors to produce electricity.

Auxiliary Beam Stress Improved Laser Welding for Repair of Irradiated Light Water Reactor Components

2019 ◽  
Author(s):  
Jian Chen ◽  
Jonathan Tatman ◽  
Zhili Feng ◽  
Roger Miller ◽  
Wei Tang ◽  
...  
Keyword(s):  
Laser Welding ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Department Of Energy ◽  
Light Water Reactor ◽  
Development Effort ◽  
Light Water ◽  
Water Reactor ◽  
Oak Ridge

Abstract The welding task focuses on development of advanced welding technologies for repair and maintenance of nuclear reactor structural components to safely and cost-effectively extend the service life of nuclear power reactors. This paper presents an integrated research and development effort by the Department of Energy Light Water Reactor Sustainability Program through the Oak Ridge National Laboratory (ORNL) and Electric Power Research Institute (EPRI) to develop a patent-pending technology, Auxiliary Beam Stress Improved Laser Welding Technique, that proactively manages the stresses during laser repair welding of highly irradiated reactor internals without helium induced cracking (HeIC). Finite element numerical simulations and in-situ temperature and strain experimental validation have been utilized to identify candidate welding conditions to achieve significant stress compression near the weld pool during cooling. Preliminary welding experiments were performed on irradiated stainless-steel plates (Type 304L). Post-weld characterization reveals that no macroscopic HeIC was observed.

Machining Test Specimens From Harvested Zion RPV Segments

2015 ◽  
Author(s):  
Thomas M. Rosseel ◽  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
National Laboratory ◽  
Compact Tension ◽  
Department Of Energy ◽  
Oak Ridge ◽  
Damage Models ◽  
The Us ◽  
Machining Test

The decommissioning of the Zion Nuclear Generating Station (NGS) in Zion, Illinois, presents a special and timely opportunity for developing a better understanding of materials degradation and other issues associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the Department of Energy (DOE), Light Water Reactor Sustainability (LWRS) Program, is coordinating and contracting with Zion Solutions, LLC, a subsidiary of Energy Solutions, an international nuclear services company, the selective procurement of materials, structures, components, and other items of interest from the decommissioned reactors. In this paper, we will discuss the acquisition of segments of the Zion Unit 2 Reactor Pressure Vessel (RPV), cutting these segments into blocks from the beltline and upper vertical welds and plate material and machining those blocks into mechanical (Charpy, compact tension, and tensile) test specimens and coupons for microstructural (TEM, SEM, APT, SANS and nano indention) characterization. Access to service-irradiated RPV welds and plate sections will allow through wall attenuation studies to be performed, which will be used to assess current radiation damage models [1].

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.

scholarly journals Current Status of the Characterization of RPV Materials Harvested From the Decommissioned Zion Unit 1 Nuclear Power Plant

2017 ◽  
Author(s):  
Thomas M. Rosseel ◽  
Mikhail A. Sokolov ◽  
Xiang Chen ◽  
Randy K. Nanstad
Keyword(s):  
Mechanical Properties ◽  
Base Metal ◽  
Nuclear Power ◽  
Power Plants ◽  
National Laboratory ◽  
Department Of Energy ◽  
Current Status ◽  
Oak Ridge ◽  
Years Of Service

The decommissioning of Units 1 and 2 of the Zion Nuclear Power Station in Zion, Illinois, after ∼ 15 effective full-power years of service presents a unique opportunity to characterize the degradation of in-service reactor pressure vessel (RPV) materials and to assess currently available models for predicting radiation embrittlement of RPV steels [1–3]. Moreover, through-wall thickness attenuation and property distributions are being obtained and the results to be compared with surveillance specimen test data. It is anticipated that these efforts will provide a better understanding of materials degradation associated with extending the lifetime of existing nuclear power plants (NPPs) beyond 60 years of service and subsequent license renewal. In support of extended service and current operations of the US nuclear reactor fleet, the Oak Ridge National Laboratory (ORNL), through the U.S. Department of Energy, Light Water Reactor Sustainability (LWRS) Program, coordinated procurement of materials, components, and other items of interest from the decommissioned Zion NPPs. In this report, harvesting, cutting sample blocks, machining test specimens, test plans, and the current status of materials characterization of the RPV from the decommissioned Zion NPP Unit 1 will be discussed. The primary foci are the circumferential, Linde 80 flux, wire heat 72105 (WF-70) beltline weld and the A533B base metal from the intermediate shell harvested from a region of peak fluence (0.7 × 1019 n/cm2, E > 1.0 MeV) on the internal surface of the Zion Unit 1 vessel. Following the determination of the through-thickness chemical composition, Charpy impact, fracture toughness, tensile, and hardness testing are being performed to characterize the through-thickness mechanical properties of base metal and beltline-weld materials. In addition to mechanical properties, microstructural characterizations are being performed using various microstructural techniques, including Atom Probe Tomography, Small Angle Neutron Scattering, and Positron Annihilation Spectroscopy.

Characterization of Advanced Steels as Accident Tolerant Cladding for Light Water Reactor Nuclear Fuel

2015 ◽  
Author(s):  
Raul B. Rebak
Keyword(s):  
Zirconium Alloy ◽  
Fission Products ◽  
Alloy System ◽  
Department Of Energy ◽  
Normal Operation ◽  
Operation Conditions ◽  
Time Period ◽  
Environmental Test ◽  
Near Term ◽  
Accident Conditions

After the tsunami incident in Fukushima in March 2011, the international community is set to identify appropriate nuclear materials with increased accident tolerance with respect to the traditional UO2-zirconium alloy fuel system permitting loss of active cooling for a considerably longer time period, while maintaining or improving the fuel performance during normal operations. The researched safety characteristics of these advanced fuels are mainly: (a) Improved reaction kinetics with steam; (b) Slower hydrogen production rate; and (c) Enhanced retention of fission products. In the US the Department of Energy is supporting the development of an improved cladding using advanced steels such as the iron-chromium-aluminum (FeCrAl) alloy system. Environmental test results show that FeCrAl alloys are highly resistant to corrosion and environmental cracking under normal operation conditions and extremely resistant to attack by steam under accident conditions. That is, the replacement of a zirconium alloy using a ferritic material containing chromium and aluminum appears to be the most near term implementation for accident tolerant fuels.

Effect of Creep on the Residual Stresses in Tube-to-Tubesheet Joints

2010 ◽  
Author(s):  
Nor Eddine Laghzale ◽  
Abdel-Hakim Bouzid
Keyword(s):  
Contact Pressure ◽  
Analytical Model ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Environmental Safety ◽  
Maintenance Cost ◽  
Joint Interface ◽  
Plant Safety ◽  
Material Creep ◽  
Nuclear Power Plant Safety

Steam generators are the subject of major concern in nuclear power plant safety. Within these generators, in addition to the structural integrity, the gross tightness barrier, which separates the primary and secondary circuits, is primarily ensured by the presence of a residual contact pressure at the tube-to-tubesheet joint interface. Any leakage is unacceptable, and its consequences are very heavy in terms of the human and environmental safety as well as maintenance cost. Some studies have been conducted to understand the main reasons for such a failure. However, no analytical model able to predict the attenuation of the residual contact pressure under the effect of material creep relaxation behavior. The development of a simple analytical model able to predict the change of the residual contact pressure as a function of time is laid out in this paper. The results from the analytical model are checked and compared with those of finite elements.

Advanced Research and Technology Development Fossil Energy Materials Program

1988 ◽  
Vol 110 (4) ◽  
pp. 670-676
Author(s):  
R. R. Judkins ◽  
R. A. Bradley
Keyword(s):  
Technology Development ◽  
National Laboratory ◽  
Development Program ◽  
Ceramic Composites ◽  
Department Of Energy ◽  
Energy Materials ◽  
Alloy Development ◽  
Fossil Energy ◽  
Oak Ridge ◽  
Research Activities

The Advanced Research and Technology Development (AR&TD) Fossil Energy Materials Program is a multifaceted materials research and development program sponsored by the Office of Fossil Energy of the U.S. Department of Energy. The program is administered by the Office of Technical Coordination. In 1979, the Office of Fossil Energy assigned responsibilities for this program to the DOE Oak Ridge Operations Office (ORO) as the lead field office and Oak Ridge National Laboratory (ORNL) as the lead national laboratory. Technical activities on the program are divided into three research thrust areas: structural ceramic composites, alloy development and mechanical properties, and corrosion and erosion of alloys. In addition, assessments and technology transfer are included in a fourth thrust area. This paper provides information on the structure of the program and summarizes some of the major research activities.

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