The effect of carbon crystal structure on treat reactor physics calculations
The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore, the carbon crystal structure is a complex mixture of graphite particles in a nongraphitized elemental carbon matrix. Results of calculations using macroscopic carbon cross sections obtained by mixing bound-kernel graphite cross sections for the graphitized carbon and free-gas carbon cross sections for the remainder of the carbon and calculations using only bound-kernel graphite cross sections are compared to experimental data. It is shown that the use of the hybridized cross sections which reflect the allotropic mixture of the carbon in the TREAT fuel results in a significant improvement in the accuracy of calculated neutronics parameters for the TREAT reactor. 6 refs., 2 figs., 3 tabs.
- Research Organization:
- Argonne National Lab., Idaho Falls, ID (USA)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 6677056
- Report Number(s):
- CONF-880911-23; ON: DE89003625
- Resource Relation:
- Conference: International reactor physics conference, Jackson Hole, WY, USA, 18 Sep 1988; Other Information: Portions of this document are illegible in microfiche products
- Country of Publication:
- United States
- Language:
- English
Similar Records
Neutronic Consideration of TREAT Facility Fuel SiC Recladding
Neutronics Analyses of the Minimum Original HEU TREAT Core
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ALLOY NUCLEAR FUELS
TESTING
CARBON
EXPERIMENTAL DATA
GRAPHITE
NEUTRONIC DAMAGE FUNCTIONS
REACTOR ACCIDENTS
REACTOR EXPERIMENTAL FACILITIES
REACTOR PHYSICS
TREAT REACTOR
URANIUM OXIDES U3O8
ACCIDENTS
ACTINIDE COMPOUNDS
AIR COOLED REACTORS
CHALCOGENIDES
DATA
ELEMENTAL MINERALS
ELEMENTS
ENERGY SOURCES
ENRICHED URANIUM REACTORS
EXPERIMENTAL REACTORS
FUELS
FUNCTIONS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HOMOGENEOUS REACTORS
INFORMATION
MATERIALS
MINERALS
NONMETALS
NUCLEAR FUELS
NUMERICAL DATA
OXIDES
OXYGEN COMPOUNDS
PHYSICS
REACTOR COMPONENTS
REACTOR MATERIALS
REACTORS
RESEARCH AND TEST REACTORS
SOLID FUELS
SOLID HOMOGENEOUS REACTORS
TEST REACTORS
THERMAL REACTORS
URANIUM COMPOUNDS
URANIUM OXIDES
050700* - Nuclear Fuels- Fuels Production & Properties
220600 - Nuclear Reactor Technology- Research
Test & Experimental Reactors