Three-dimensional thermo-fluiddynamic analysis of gas flow in straight piping with WINDFLOW code
Creators
- 1. Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Dept. of Reactor Safety Research
- 2. Mitsubishi Research Inst., Inc., Tokyo (Japan). Research Center for Safety Science
Description
During a postulated severe accident of a light water reactor, the temperature of core component materials is elevated due to a loss of cooling capability, resulting in formation of a degraded core. Fission products (FPs) are released from the degraded core during a process of the accident progression. Most of FPs could be transported within a reactor coolant system with a high temperature gaseous flow as a form of aerosol. Deposition of FP aerosols onto the inner surface of the reactor piping could be realized by several physical and chemical mechanisms. A thermal load caused by decay heat emission from the deposited aerosols is imposed on the piping structure. Additional thermal load onto the piping is thought to be resulted from a convective flow of high temperature gases form the degraded core region. The integrity of the reactor piping system, therefore, is challenged by these thermal loads, in addition to a pressure load in some accident sequences. A computer code, WINDFLOW, for a three-dimensional thermo-fluiddynamic analysis of a single phase gaseous flow in a reactor coolant piping has been developed at JAERI. A hybrid grid system composed of triangular and quadrangular cross-sectional cells was introduced in WINDFLOW to enhance the modeling flexibility. WINDFLOW was applied to an analysis of tests on thermo-fluiddynamics within a piping performed in WIND project. Formation of a cross-sectional secondary flow resulted from the natural convection was predicted. Appearance of a sharp cross-sectional temperature gradient at the vicinity of the piping wall was confirmed in the analysis. The predictions on the gaseous temperature distribution were compared with the test results. It was found that WINDFLOW tended to underestimate the axial temperature decrease and could qualitatively well reproduce radial temperature distribution observed in the tests
Additional details
Publishing Information
- Publisher
- American Society of Mechanical Engineers.
- Imprint Place
- New York, NY (United States)
- ISBN
- 0-7918-1226-X
- Imprint Title
- ICONE-4: Proceedings. Volume 1 -- Part B: Basic technological advances
- Imprint Pagination
- 564 p.
- Journal Page Range
- p. 997-1008.
Conference
- Title
- ASME/JSME international conference on nuclear engineering.
- Acronym
- ICONE 4
- Dates
- 10-13 Mar 1996.
- Place
- New Orleans, LA (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 27068612
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Numerical Data
- Descriptors DEI
- ENVIRONMENTAL TRANSPORT; FISSION PRODUCT RELEASE; GAS FLOW; HEAT TRANSFER; HYDRAULICS; NUMERICAL ANALYSIS; PIPES; RADIOACTIVE AEROSOLS; REACTOR ACCIDENTS; REACTOR COOLING SYSTEMS; TEMPERATURE DISTRIBUTION; THEORETICAL DATA; THREE-DIMENSIONAL CALCULATIONS; W CODES; WATER COOLED REACTORS
- Descriptors DEC
- ACCIDENTS; AEROSOLS; COLLOIDS; COMPUTER CODES; COOLING SYSTEMS; DATA; DISPERSIONS; ENERGY TRANSFER; FLUID FLOW; INFORMATION; MASS TRANSFER; MATHEMATICS; NUMERICAL DATA; REACTOR COMPONENTS; REACTORS; SOLS
Optional Information
- Secondary number(s)
- CONF-960306--.