Update on Overcoating and Compacting Activities for Coated Particles With 425 µm Kernels

2008 ◽  
Author(s):  
Peter J. Pappano ◽  
John D. Hunn
Keyword(s):  
Circular Cylinder ◽  
Uranium Oxide ◽  
Nuclear Plant ◽  
Next Generation ◽  
Coated Particles ◽  
Uranium Carbide ◽  
Test Reactor

The Advanced Gas Reactor (AGR) program is tasked with developing and qualifying fuel for the Next Generation Nuclear Plant (NGNP) [1, 2]. The first experiment, AGR-1, focused on TRISO coating 350 μm uranium oxide/uranium carbide (UCO) kernels and compacting them into a right circular cylinder fuel form using an overcoating and compacting process. The AGR-1 fuel compacts are currently being irradiated at the Advanced Test Reactor (ATR). The AGR-2 experiment will focus on overcoating and compacting TRISO coated 425 μm UCO kernels. This paper summaries the work that has been done to date on preparing to make AGR-2 compacts.

Thermal Evaluations for Next Generation Nuclear Plant Fuel Testing

Volume 4 ◽  
2004 ◽  
Author(s):  
Richard G. Ambrosek ◽  
Debbie J. Utterbeck ◽  
Brandon Miller
Keyword(s):  
Nuclear Reactor ◽  
Fuel Cycle ◽  
Nuclear Plant ◽  
Light Water Reactors ◽  
Next Generation ◽  
Preliminary Information ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Test Reactor ◽  
Test Fuel

The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility.

Mission and Status of the First Two Next Generation Nuclear Plant Fuel Irradiation Experiments in the Advanced Test Reactor

2010 ◽  
Author(s):  
S. Blaine Grover ◽  
David A. Petti ◽  
John T. Maki
Keyword(s):  
United States ◽  
Fission Product ◽  
The United States ◽  
Nuclear Plant ◽  
United States Department ◽  
Next Generation ◽  
States Department ◽  
On Line ◽  
Test Reactor ◽  
And Control

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to nine low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and the irradiations will be completed over the next five to six years to support demonstration and qualification of new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of multiple separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) is currently being fabricated and assembled for insertion in the ATR in the early to mid calendar 2010. The design of test trains, the support systems and the fission product monitoring system used to monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the first two experiments will be compared, and updated information on the design and status of AGR-2 is provided. The preliminary irradiation results for the AGR-1 experiment are also presented.

scholarly journals Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

2008 ◽  
Author(s):  
Richard R. Schultz ◽  
Abderrafi M. Ougouag ◽  
David W. Nigg ◽  
Hans D. Gougar ◽  
Richard W. Johnson ◽  
...  
Keyword(s):  
Research And Development ◽  
Nuclear Plant ◽  
Next Generation ◽  
Technical Program ◽  
Program Plan

Project Deep-Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors

2008 ◽  
Author(s):  
Robert M. Versluis ◽  
Francesco Venneri ◽  
David Petti ◽  
Lance Snead ◽  
Donald McEachern
Keyword(s):  
High Temperature ◽  
Nuclear Plant ◽  
Next Generation ◽  
Very High Temperature Reactor ◽  
High Temperature Reactor ◽  
Very High ◽  
The U.S

The helium-cooled, graphite-moderated Very High Temperature Reactor (VHTR) has become the centerpiece of the U.S. Department of Energy’s (DOE) Next Generation Nuclear Plant (NGNP) program. The NGNP program aims to construct a VHTR prototype, with the participation of industry, by the year 2021.

A Brief Review of Models Representing Creep of Alloy 617

2005 ◽  
Author(s):  
Robert W. Swindeman ◽  
Michael J. Swindeman ◽  
Weiju Ren
Keyword(s):  
High Temperature ◽  
Nuclear Plant ◽  
Next Generation ◽  
Alloy 617 ◽  
Temperature Range ◽  
Damage Models ◽  
The Us ◽  
Creep Models ◽  
High Temperature Gas

Alloy 617 is being considered for the construction of components to operate in the Next Generation Nuclear Plant (NGNP). Service temperatures will range from 650 to 1000°C. To meet the needs of the conceptual designers of this plant, a materials handbook is being developed that will provide information on alloy 617, as well as other materials of interest. The database for alloy 617 to be incorporated into the handbook was produced in the 1970s and 1980s, while creep and damage models were developed from the database for use in the design of high-temperature gas-cooled reactors. In the work reported here, the US database and creep models are briefly reviewed. The work reported represents progress toward a useful model of the behavior of this material in the temperature range of 650 to 1000°C.

scholarly journals Next Generation Nuclear Plant Methods Technical Program Plan

2010 ◽  
Author(s):  
Richard R. Schultz ◽  
Abderrafi M. Ougouag ◽  
David W. Nigg ◽  
Hans D. Gougar ◽  
Richard W. Johnson ◽  
...  
Keyword(s):  
Nuclear Plant ◽  
Next Generation ◽  
Technical Program ◽  
Program Plan
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