An Integral Effect Test of Complete Loss of RCS Flowrate for SMART with the VISTA-ITL
- 1. Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
Description
A thermal-hydraulic integral effect test facility, VISTA-ITL, for the SMART design has been constructed by the Korea Atomic Energy Research Institute (KAERI). The VISTA-ITL is a revised version of an existing VISTA facility to have the simulation capability of small-break loss of coolant accident (SBLOCA), passive residual heat removal system (PRHRS) performance tests, and complete loss of RCS (Reactor Coolant System) flowrate (CLOF), etc. It is newly equipped with the steam pressurizer, the safety injection system, the steam generator bypass, the pump discharge line, the dowcomer, the PRHRS makeup tank, the break simulator and the break measuring system, etc. The VISTA-ITL is a 1/2.77-height, 1/1310-volume scaled test facility based on the design features of SMART. The reference scale ratios of length (1/2.77) and area (1/472.9) are based on the elevation difference between the core and steam generator centers and the core flow area, respectively. According to the scaling law, the reduced height scaling results in time-reducing results in the model as the time scale ratio is 1/1.664. A CLOF test was successfully performed by using the VISTA-ITL facility recently and its major results have been discussed
Additional details
Publishing Information
- Publisher
- KNS
- Imprint Place
- Daejeon (Korea, Republic of)
- Imprint Title
- Proceedings of the KNS autumn meeting
- Imprint Pagination
- [1 CD-ROM]
- Journal Page Range
- [2 p.]
Conference
- Title
- 2011 autumn meeting of the KNS
- Dates
- 26-28 Oct 2011
- Place
- Kyoungju (Korea, Republic of)
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 43045472
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- COMPUTERIZED SIMULATION; DESIGN; FLOW RATE; LOSS OF COOLANT; PWR TYPE REACTORS; TEST FACILITIES; THERMAL HYDRAULICS
- Descriptors DEC
- ACCIDENTS; ENRICHED URANIUM REACTORS; FLUID MECHANICS; HYDRAULICS; MECHANICS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 5 refs, 1 fig, 3 tabs