Published 1999
| Version v1
Book
ASSERT-PV simulations of two-phase flow in horizontal and vertical subchannels
Creators
- 1. Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)
Description
This is a part of the effort to assess the ASSERT-PV code which is supposedly capable of quantifying the effect of small flow boundary changes in the fuel channel of CANDU reactors. Two independently performed subchannel experiments are simulated by the ASSERT-PV code. The result includes the pressure and the void fraction distributions in each subchannel. It is found that the ASSERT-PV predicts both experimental data quite well by selecting the void diffusion constant properly for the adiabatic two-phase flows. (author)
Availability note (English)
Available on CD-ROM from Canadian Nuclear Association, Toronto, Ontario, Canada M5J 2L7Additional details
Publishing Information
- Publisher
- Canadian Nuclear Association, Canadian Nuclear Society
- Imprint Place
- Toronto, Ontario (Canada)
- Imprint Title
- Proceedings of the 39. annual Canadian Nuclear Association conference and 20. annual Canadian Nuclear Society conference
- Imprint Pagination
- 640 Megabytes
- Journal Page Range
- (Session A2) [6 p.]
- ISSN
- 0227-1907
Conference
- Title
- Communicating the nuclear advantage
- Dates
- 30 May - 2 Jun 1999
- Place
- Montreal, PQ (Canada)
INIS
- Country of Publication
- Canada
- Country of Input or Organization
- Canada
- INIS RN
- 32002703
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- A CODES; CANDU TYPE REACTORS; FUEL CHANNELS; FUEL ELEMENT CLUSTERS; TWO-PHASE FLOW; VOID COEFFICIENT
- Descriptors DEC
- COMPUTER CODES; FLUID FLOW; FUEL ASSEMBLIES; HEAVY WATER MODERATED REACTORS; POWER REACTORS; PRESSURE TUBE REACTORS; REACTIVITY COEFFICIENTS; REACTOR CHANNELS; REACTOR COMPONENTS; REACTORS; THERMAL REACTORS
Optional Information
- Notes
- 12 refs., 1 tab., 7 figs.