An assessment of the best estimate thermal-hydraulic analysis code CATHARE on CREARE downcomer experiment
- 1. Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
Description
A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical loss of coolant accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate, ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically. (Author)
Additional details
Publishing Information
- Journal Title
- Journal of the Korean Nuclear Society
- Journal Volume
- 24
- Journal Issue
- 3
- Journal Page Range
- p. 274-284.
- ISSN
- 0372-7327
- CODEN
- WJHKAW
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 24036083
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S99: GENERAL AND MISCELLANEOUS;
- Descriptors DEI
- COMPUTER CODES; COOLING SYSTEMS; COUNTER CURRENT; FLUID FLOW; HYDRAULICS; LOSS OF COOLANT; PWR TYPE REACTORS; SIMULATION; STEAM LINES; THERMODYNAMICS
- Descriptors DEC
- ACCIDENTS; ENRICHED URANIUM REACTORS; PIPELINES; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS