Analysis of core flow deviation caused by the large gaps between horizontal block interfaces
- 1. Department of Power Reactor Projects, Japan Atomic Energy Research Inst., Tokai, Ibaraki
Description
There are various leak flow paths between blocks in the experimental VHTR core. Especially, leak flow in the gap between horizontal surfaces of fuel blocks is called ''crossflow''. Generally, the coolant flows from the fuel channel into the gap between columns. As its results fuel temperature rises. In case of the birth of large gap between horizontal block interfaces, the crossflow caused by it effects significantly on the core flow distribution, core temperature distribution and F.P. release. In this paper, the horizontal core offset between the core upper structure and the core support structure is postulated as the cause of occurrence of large horizontal gap at fuel column. A prediction method of large horizontal gaps between blocks, and analytical method and results of core flow distribution are discussed. Analytical results of core behavior in the core flow deviation caused by the gaps between horizontal block interfaces are also presented. Evaluation of fission product release depending on the flow distribution change is also carried out. (author)
Additional details
Publishing Information
- Imprint Title
- Gas-cooled reactor safety and accident analysis
- Imprint Pagination
- 349 p.
- Journal Page Range
- p. 313-323.
- Report number
- IAEA-TECDOC--358
Conference
- Title
- Specialists' meeting on safety and accident analysis for gas-cooled reactors.
- Dates
- 13-15 May 1985.
- Place
- Oak Ridge, TN (USA).
INIS
- Country of Publication
- International Atomic Energy Agency (IAEA)
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 17036739
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPUTERIZED SIMULATION; COOLANTS; FISSION PRODUCT RELEASE; FUEL CHANNELS; GAS COOLING; GAS FLOW; HIGH TEMPERATURE; PRIMARY COOLANT CIRCUITS; TEMPERATURE DISTRIBUTION; VHTR REACTOR
- Descriptors DEC
- COOLING; COOLING SYSTEMS; ENRICHED URANIUM REACTORS; EXPERIMENTAL REACTORS; FLUID FLOW; GAS COOLED REACTORS; GRAPHITE MODERATED REACTORS; HELIUM COOLED REACTORS; HTGR TYPE REACTORS; POWER REACTORS; REACTOR CHANNELS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; RESEARCH AND TEST REACTORS; SIMULATION; THERMAL REACTORS
Optional Information
- Notes
- 5 refs, 9 figs, 3 tabs.