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.

scholarly journals Concept of Erbium Doped Uranium Oxide Fuel Cycle in Light Water Reactors

2002 ◽  
Vol 39 (5) ◽  
pp. 506-513 ◽  
Author(s):  
Vladimir BARCHEVTSEV ◽  
Vladimir ARTISYUK ◽  
Hisashi NINOKATA
Keyword(s):  
Uranium Oxide ◽  
Fuel Cycle ◽  
Oxide Fuel ◽  
Light Water Reactors ◽  
Light Water ◽  
Erbium Doped ◽  
Water Reactors

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.

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.

Methodology for the Determination of a Set of Safety Factors That are Consistent for Design Code and Fitness-for-Service Code: Framework

2009 ◽  
Author(s):  
Tai Asayama
Keyword(s):  
Reliability Evaluation ◽  
Light Water Reactors ◽  
Next Generation ◽  
Safety Factors ◽  
Design Code ◽  
Reliability Indices ◽  
Allowable Stresses ◽  
Complete Set ◽  
Water Reactors

This paper introduces a methodology for the determination of a complete set of safety factors that maintains consistency between design code and fitness-for-service code of nuclear components. The purpose of the work is to materialize the System Based Code concept, which is indispensable for the development of next generation nuclear reactors. The methodology consists of three principles proposed by the author which should be the basis of code development for new next generation reactors. The principles are; 1) Design to target reliability, 2) Continuous reliability evaluation from design to fitness-for-service, 3) Update of reliability evaluation based on information obtained during construction and operation. Effectiveness of the methodology is demonstrated using a simple example problem. The problem deals with pipe subjected to internal pressure under conditions which is typical in light water reactors. Following the reliability evaluation of current situation which meets the provisions of design code and fitness-for-serve code published from Japan Society of Mechanical Engineers, the three principles are applied step-by-step and safety factors and reliability indices are newly derived. It is shown that a complete application of the three principles could lead to a set of safety factors that assures consistency in terms of reliability in design and fitness-for-service, and improves allowable stresses as well. Technologies to be developed and issues to be discussed for application of the methodology to more complicated and practical situations are described as well.

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