Published May 2002
| Version v1
Miscellaneous
A three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR
- 1. Korea Maritime Univ., Pusan (Korea, Republic of)
- 2. Handong, Univ., Pohang (Korea, Republic of)
- 3. KINS, Taejon (Korea, Republic of)
Description
Three-dimensional analyses of fluid flow and heat transfer has been performed in this study. The simulation of SPEL experimental work and comparison with experimental data has been carried out to verify the analyses models. Moreover, to verify the CANDU-6 reactor type, analyses of fluid flow and heat transfer in the calandria under the condition of steady state has been performed using FLUENT code, which is the conventional code for a three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR thermal-hydraulics. It is found that the maximum temperature in the moderator is 347K (74 ), so that the moderator has the enough subcoolability to ensure the integrity of pressure tube during LOCA conditions
Additional details
Publishing Information
- Publisher
- KNS
- Imprint Place
- Taejon (Korea, Republic of)
- Imprint Title
- Proceedings of the KNS-KARP Joint spring meeting
- Imprint Pagination
- [CD-ROM]
- Journal Page Range
- [12 p.]
Conference
- Title
- 2002 joint spring meeting of the KNS-KARP
- Dates
- 23-24 May 2002
- Place
- Gwangju (Korea, Republic of)
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 34009716
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Numerical Data, Non-conventional Literature
- Descriptors DEI
- EXPERIMENTAL DATA; F CODES; FLUID FLOW; HEAT TRANSFER; PHWR TYPE REACTORS; PRESSURE TUBES
- Descriptors DEC
- COMPUTER CODES; DATA; ENERGY TRANSFER; HEAVY WATER COOLED REACTORS; HEAVY WATER MODERATED REACTORS; INFORMATION; NUMERICAL DATA; REACTORS; TUBES
Optional Information
- Notes
- 5 refs, 16 figs, 2 tabs