Analysis of void fraction in single channel using TRACE, MARS-KS, and RELAP5
Creators
- 1. KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of)
Description
The accurate prediction of the void fraction is one of the most important factors in subchannel analyses. In general, the subchannel analysis has been conducted by means of dedicated subchannel analysis codes. However, the recent development of system codes with advanced two-phase flow model and three-dimensional components has helped more accurate and precise prediction of multiphase phenomena in nuclear reactor systems. This paper aims at evaluating the applicability of three different system codes to the prediction of the void fraction based on NUPEC experiment results employed for OECD/NRC PSBT benchmark. As a first step to the full scope analysis, analysis results for single channel experiments by using TRACE 5.0 Patch 3, MARS-KS 1.3, and RELAP5/MOD3.3 Patch 4 are presented in this paper. The result indicates that all codes slightly over-predict the void fraction compared to the experimental results in general and no significant discrepancies between the codes are observed. (author)
Additional details
Publishing Information
- Imprint Title
- Proceedings of the 10th international topical meeting on nuclear thermal hydraulics, operation and safety (NUTHOS-10)
- Imprint Pagination
- 2846 p.
- Journal Page Range
- 10 p.
Conference
- Title
- 10. international topical meeting on nuclear thermal hydraulics, operation and safety
- Acronym
- NUTHOS-10
- Dates
- 14-18 Dec 2014
- Place
- Ginowan, Okinawa (Japan)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 47077874
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- COMPUTER CODES; DEPARTURE NUCLEATE BOILING; FORECASTING; FUEL ASSEMBLIES; NEUTRON PHYSICS; NUCLEAR POWER PLANTS; PRESSURE RANGE MEGA PA 10-100; PWR TYPE REACTORS; REACTOR SAFETY; REGRESSION ANALYSIS; TEMPERATURE RANGE 0400-1000 K; THERMAL HYDRAULICS; TRANSIENTS; TWO-PHASE FLOW; VOID FRACTION
- Descriptors DEC
- BOILING; ENRICHED URANIUM REACTORS; FLUID FLOW; FLUID MECHANICS; HYDRAULICS; MATHEMATICS; MECHANICS; NUCLEAR FACILITIES; NUCLEATE BOILING; PHASE TRANSFORMATIONS; PHYSICS; POWER PLANTS; POWER REACTORS; PRESSURE RANGE; PRESSURE RANGE MEGA PA; REACTORS; SAFETY; STATISTICS; TEMPERATURE RANGE; THERMAL POWER PLANTS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- Available as USB Flash Memory Data in PDF format. Paper ID: NUTHOS10-1089.pdf; 8 refs., 7 figs., 4 tabs.