In 2000, British Nuclear Fuels Limited (BNFL) commissioned an irradiation program at the United States Department of Energy’s Idaho National Engineering and Environmental Laboratory (INEEL) to assess the effects of extended operating scenarios upon the integrity of Magnox reactor cores. In this program, predictions of thermal and physical effects on these graphite cores were developed using analytical computer models. To benchmark results, experimental graphite assemblies representative of the Magnox graphite were irradiated in the Advanced Test Reactor (ATR). This paper analyzes and contrasts the thermal predictions with those experimental results. These investigations were conducted to extend existing graphite physical property databases for higher radiolytic weight loss (35–50% density reduction) than occur during the economic planning life of these reactors. These data then can be used to make extended life projections regarding the suitable function of the graphite in its various roles of providing the physical structure for the fuel, neutron moderator, medium for instrumentation, and coolant channels. Extended irradiation effects will be obtained with samples of archived, pre-characterized graphite used in the Magnox type reactors. The new Irradiation Test Vehicle (ITV) facility in the ATR contained the experiments and provided the desired irradiation conditions as well as on-line temperature control. The capability to provide both oxidizing and inert gas atmospheres for the graphite specimens was added to the ITV to enable assessment of the individual and combined effects of oxidation and neutron damage to the specimens. In this paper the thermal evaluations (performed to size the control gaps to obtain the desired thermal performance) are contrasted to actual experimental results.
Skip Nav Destination
ASME 2003 Heat Transfer Summer Conference
July 21–23, 2003
Las Vegas, Nevada, USA
Conference Sponsors:
- Heat Transfer Division
ISBN:
0-7918-3694-0
PROCEEDINGS PAPER
Comparison of Predictive and Experimental Data From Graphite Irradiations in the Advanced Test Reactor Irradiation Test Vehicle
Richard G. Ambrosek,
Richard G. Ambrosek
Idaho National Engineering and Environmental Laboratory (INEEL), Idaho Falls, ID
Search for other works by this author on:
Debbie J. Utterbeck
Debbie J. Utterbeck
Idaho National Engineering and Environmental Laboratory (INEEL), Idaho Falls, ID
Search for other works by this author on:
Richard G. Ambrosek
Idaho National Engineering and Environmental Laboratory (INEEL), Idaho Falls, ID
Debbie J. Utterbeck
Idaho National Engineering and Environmental Laboratory (INEEL), Idaho Falls, ID
Paper No:
HT2003-47396, pp. 531-535; 5 pages
Published Online:
December 17, 2008
Citation
Ambrosek, RG, & Utterbeck, DJ. "Comparison of Predictive and Experimental Data From Graphite Irradiations in the Advanced Test Reactor Irradiation Test Vehicle." Proceedings of the ASME 2003 Heat Transfer Summer Conference. Heat Transfer: Volume 2. Las Vegas, Nevada, USA. July 21–23, 2003. pp. 531-535. ASME. https://doi.org/10.1115/HT2003-47396
Download citation file:
8
Views
Related Proceedings Papers
Related Articles
Elaboration of Conductive Thermal Storage Composites Made of Phase Change Materials and Graphite for Solar Plant
J. Sol. Energy Eng (February,2008)
Calculation of the Response of a Composite Plate to Localized Dynamic Surface Loads Using a New Wave Number Integral Method
J. Appl. Mech (January,2005)
Cobalt Sulphate as an Alternative Counter Electrode Material in Dye Sensitized Solar Cells
J. Sol. Energy Eng (November,2014)
Related Chapters
Nonmetallic Pressure Piping System Components Part A: Experience With Nonmetallic Materials in Structural/Pressure Boundary Applications
Online Companion Guide to the ASME Boiler & Pressure Vessel Codes
Division 5—High Temperature Reactors
Online Companion Guide to the ASME Boiler & Pressure Vessel Codes
Division 5—High Temperature Reactors
Companion Guide to the ASME Boiler and Pressure Vessel Codes, Volume 1, Fifth Edition