A study of neutron diffusion in sintered beryllium oxide by a pulse method

Journal Article · Sun Dec 01 04:00:00 EST 1963 · Soviet Atomic Energy

DOI:https://doi.org/10.1007/BF01115834· OSTI ID:4704530

Zhezherun, I. F.

BeO, which is an excellent moderator material, possesses high heat resistance, has low absorption cross section, and provides for supplementary neutron multiplication by the Be/sup 9/(n,2n) reaction, is considered a very promising material for high-temperature power reactors. In view of the contradictory data reported in the literature concerning the neutron-diffusion parameters in the sintered material, further thermal neutron-diffusion studies were undertaken using the pulsed-source method. A specimen previously used for determining the slowing-down length (Atomnaya Energ., 13: 258(1962)) was employed in this study. The values of the diffusion coefficient, diffusion length, transport length, effective transport cross section, absorption cross section, and the diffusion time were calculated and plotted. The calculated time of thermalization, 120 mu sec, was considerably lower than the value reported in the literature; this might be due to the approximating assumptions used. (TTT)

• Research Organization: Originating Research Org. not identified
• Sponsoring Organization: USDOE
• NSA Number: NSA-17-024374
• OSTI ID: 4704530
• Journal Information: Soviet Atomic Energy, Journal Name: Soviet Atomic Energy Journal Issue: 2 Vol. 14; ISSN 0038-531X
• Country of Publication: Country unknown/Code not available
• Language: Russian

Related Subjects

ABSORPTION
BERYLLIUM 9
BERYLLIUM OXIDES
CROSS SECTIONS
DIFFUSION
DIFFUSION LENGTH
HEATING
MODERATORS
NEUTRON SOURCES
NEUTRONS
NUCLEAR REACTIONS
NUMERICALS
PHYSICS
PULSES
REACTORS
SINTERED MATERIALS
SLOWDOWN
TEMPERATURE
THERMAL NEUTRONS
TRANSPORT THEORY