Experiments on the heat transfer and natural circulation characteristics of the passive residual heat removal system for an advanced integral type reactor
Creators
- 1. Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)
Description
Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS. (author)
Additional details
Identifiers
Publishing Information
- Journal Title
- Journal of Nuclear Science and Technology (Tokyo)
- Journal Volume
- 44
- Journal Issue
- 5
- Journal Page Range
- p. 703-713
- ISSN
- 0022-3131
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 53076893
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- AFTER-HEAT REMOVAL; COOLANT LOOPS; HEAT EXCHANGERS; KAERI; MOCKUP; MODULAR STRUCTURES; NATURAL CONVECTION; PERFORMANCE; PWR TYPE REACTORS; REACTOR ACCIDENTS; TEST FACILITIES; THERMAL HYDRAULICS; TRANSIENTS; TWO-PHASE FLOW
- Descriptors DEC
- ACCIDENTS; CONVECTION; COOLING SYSTEMS; ENERGY SYSTEMS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUID FLOW; FLUID MECHANICS; HEAT TRANSFER; HYDRAULICS; KOREAN ORGANIZATIONS; MASS TRANSFER; MECHANICS; NATIONAL ORGANIZATIONS; POWER REACTORS; REACTORS; REMOVAL; STRUCTURAL MODELS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 11 refs., 19 figs., 3 tabs.; This record replaces 38107742