Influence of creep anisotropy on the stress-strain state of channel tubes

Journal Article · Mon Feb 01 04:00:00 EST 1988 · Sov. At. Energy (Engl. Transl.); (United States)

OSTI ID:6911105

Zheltukhin, K K; Tutnov, A A; Ul'yanov, A I

The authors compile a mathematical model depicting creep behavior and swelling under the effects of neutron irradiation and temperature in zirconium alloy reactor channels. The model accounts for coolant pressure and thermoradiational creep anisotropy and performs a stress-strain analysis which incorporates elastic and plastic deformation and possible mechanical interference with graphite moderators. Test results against experimental conditions encountered in the RBMK-1000 reactor are given.

• OSTI ID: 6911105
• Journal Information: Sov. At. Energy (Engl. Transl.); (United States), Journal Name: Sov. At. Energy (Engl. Transl.); (United States) Vol. 63:2; ISSN SATEA
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210300 -- Power Reactors
Nonbreeding
Graphite Moderated
36 MATERIALS SCIENCE
360103 -- Metals & Alloys-- Mechanical Properties
360106* -- Metals & Alloys-- Radiation Effects
ALLOYS
ANISOTROPY
CARBON
CREEP
DEFORMATION
ELASTICITY
ELEMENTAL MINERALS
ELEMENTS
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUID MECHANICS
GRAPHITE
GRAPHITE MODERATED REACTORS
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
LENINGRAD-1 REACTOR
LWGR TYPE REACTORS
MATHEMATICAL MODELS
MECHANICAL PROPERTIES
MECHANICS
MINERALS
NEUTRON FLUX
NONMETALS
PHYSICAL RADIATION EFFECTS
PLASTICITY
POWER REACTORS
PRESSURE EFFECTS
RADIATION EFFECTS
RADIATION FLUX
REACTOR CHANNELS
REACTOR COMPONENTS
REACTORS
STRAINS
STRESS ANALYSIS
STRESSES
SWELLING
TEMPERATURE EFFECTS
TENSILE PROPERTIES
THERMAL REACTORS
TUBES
WATER COOLED REACTORS
ZIRCONIUM ALLOYS