Published January 2006 | Version v1
Journal article Metadata-only

Safety demonstration tests in HTGR (Control Rod Withdrawal Test)

Creators

  • 1. Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

Description

The HTTR (High Temperature Engineering Test Reactor), which had been developed as a helium-cooled and graphite-moderated test reactor aiming at establishment of the HTGR technology, achieved a reactor outlet coolant temperature of 950degC in April 2004. In order to upgrade the HTGR technology, control rod withdrawal tests at initial power of 30-60% rated power were conducted as safety demonstration tests in HTGR. Test results were evaluated with one channel-one temperature coefficient model and a newly developed multi-channel model using different temperature coefficients. The multi-channel model agreed better with power peak and transient behavior of tests. (T.Tanaka)

Additional details

Publishing Information

Journal Title
Koon Gakkai-Shi
Journal Volume
32
Journal Issue
1
Journal Page Range
p. 16-26
ISSN
0387-1096

INIS

Country of Publication
Japan
Country of Input or Organization
Japan
INIS RN
37056342
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Descriptors DEI
COATED FUEL PARTICLES; FUEL CHANNELS; GRAPHITE; HTTR REACTOR; REACTOR KINETICS; REACTOR SAFETY EXPERIMENTS; ROD EJECTION ACCIDENTS; TEMPERATURE COEFFICIENT; TRANSIENTS
Descriptors DEC
ACCIDENTS; CARBON; ELEMENTS; ENRICHED URANIUM REACTORS; EXPERIMENTAL REACTORS; FUEL PARTICLES; GAS COOLED REACTORS; GRAPHITE MODERATED REACTORS; HELIUM COOLED REACTORS; HTGR TYPE REACTORS; KINETICS; MINERALS; NONMETALS; REACTIVITY COEFFICIENTS; REACTOR ACCIDENTS; REACTOR CHANNELS; REACTOR COMPONENTS; REACTORS; RESEARCH AND TEST REACTORS

Optional Information

Notes
14 refs., 34 figs., 4 tabs.