MONTE CARLO RESEARCH SERIES: TWO EXPERIMENTS ON VOID GAP EFFECTS IN A ONE DIMENSIONAL HETEROGENEOUS SYSTEM
Technical Report
·
OSTI ID:4801450
The diffusion length, escape fraction and absorption point distribution were computed for a one-dimensional heterogeneous system of alternating solid and void segments. These results are compared with those computed for the homogeneous counterpart of the heterogeneous system. The solid segments were 1.0 cm long. The void segment length varied from 0.4 to 2.0 cm. The total cross section in a solid segment was 2.0 cm/sup -1/and the absorption cross section in the range 0.1--0.4 cm/sup -1/. The difference in diffusion length, between a heterogeneous and homogeneous system, was less than 7% for a point source located at the end of a solid segment and less than 1.3% for a point source located at the center of a solid segment. In a finite geometry the homogeneous model overe stimated the leakage fraction and underestimated the absorption point density by as much as 33% in a solid segment. (auth)
- Research Organization:
- General Electric Co. Aircraft Nuclear Propulsion Dept., Cincinnati
- DOE Contract Number:
- AT(11-1)-171
- NSA Number:
- NSA-16-008304
- OSTI ID:
- 4801450
- Report Number(s):
- DC-58-5-97
- Country of Publication:
- United States
- Language:
- English
Similar Records
MONTE CARLO RESEARCH SERIES: MONTE CARLO PROGRAM FOR A LINEAR REACTOR WITH VOID GAPS
DIFFUSION TENSOR FOR SLAB GEOMETRY
Verification of void effect calculation
Technical Report
·
Tue Nov 12 23:00:00 EST 1957
·
OSTI ID:4298076
DIFFUSION TENSOR FOR SLAB GEOMETRY
Technical Report
·
Sun Nov 30 23:00:00 EST 1958
·
OSTI ID:4231159
Verification of void effect calculation
Journal Article
·
Sat Dec 30 23:00:00 EST 1995
· Transactions of the American Nuclear Society
·
OSTI ID:411779
Related Subjects
ABSORPTION
AIRCRAFT
BUCKLING
CONFIGURATION
CROSS SECTIONS
DIFFUSION LENGTH
EQUATIONS
ERRORS
FUELS
HOMOGENEOUS REACTORS
LEAKS
LOSSES
MEASURED VALUES
METALS
MODERATORS
MONTE CARLO METHOD
NEUTRON FLUX
NEUTRON SOURCES
NUMERICALS
PROPULSION
REACTIVITY
REACTOR CORE
REACTOR TECHNOLOGY
REACTORS
RESONANCE ESCAPE PROBABILITY
SOLIDS
THERMAL UTILIZATION
ZONES
AIRCRAFT
BUCKLING
CONFIGURATION
CROSS SECTIONS
DIFFUSION LENGTH
EQUATIONS
ERRORS
FUELS
HOMOGENEOUS REACTORS
LEAKS
LOSSES
MEASURED VALUES
METALS
MODERATORS
MONTE CARLO METHOD
NEUTRON FLUX
NEUTRON SOURCES
NUMERICALS
PROPULSION
REACTIVITY
REACTOR CORE
REACTOR TECHNOLOGY
REACTORS
RESONANCE ESCAPE PROBABILITY
SOLIDS
THERMAL UTILIZATION
ZONES