Computational fluid dynamics analysis of buoyancy driven stratified flow
Creators
- 1. Gesellschaft fuer Anlagen- und Reaktorsicherheit GmbH, Forschungsinstitute, 85748 Garching (Germany)
Description
In the framework of the European Nuclear Reactor Integrated Simulation Project (NURISP), Computation Fluid Dynamics (CFD) software is validated for the simulation of fluid flow and heat transfer related to pressurized thermal shocks (PTS). One of the proposed validation experiments are the test series performed within the OECD ROSA V project in the Large Scale Test Facility LSTF (JAERI, 2003). The LSTF is a 1:48 volume scaled model of a four-loop Westinghouse pressurized water reactor (PWR). The ROSA V Test T1-1 investigates the temperature stratification under natural circulation conditions. Its main purpose is the validation of three-dimensional CFD calculations. The experiment was performed in several steps. It started with emergency core cooling (ECCS) injections into the cold legs at 15.5 MPa at 100 pc primary inventory with a core power corresponding to 2 pc of the scaled nominal power. In 10-minute intervals for re-stabilisation, the water level was reduced to 80 pc, 70 pc and 50 pc of the inventory. Multi-dimensional temperature distributions were measured with thermocouple rakes in the cold legs located in three cross-sectional planes between injection nozzle and downcomer. 18 thermocouples were installed below each cold leg. The nominal accuracy of the thermocouple measurements was ± 2.75 K. CFD calculations were performed for a single phase flow injection into Loop A using ANSYS CFD software. Following the OECD/NEA Best Practice Guidelines for the use of CFD in nuclear reactor safety applications (Mahaffey et al., 2007), three hexahedral grids were generated. The coarse grid had 500 000 elements. The grids were scalable with a minimum grid angle of 32 deg.. Iteration and discretization errors in time and space were checked on each grid. The transient calculations were started from a steady-state solution of the natural circulation. At the pump positions, and at ECC nozzle A, the measured mass flow rates and, in the first set of simulations, a constant inlet temperature was applied. In a second step, the measured temperature at the ECC-nozzle was used as inlet boundary condition. At the outlet, which was positioned at the lower part of the downcomer, a pressure boundary condition was prescribed. A number of calculations were performed applying symmetry boundary conditions in the downcomer. This half-model was used for checking the influence of the discretization schemes, and turbulence models (Shear Stress Transport and Reynolds stress model). The final transient calculations were obtained in a complete model of the downcomer with cold legs A and B of the LSTF facility. The results were in very good agreement with data. (authors)
Additional details
Publishing Information
- Imprint Title
- Computational Fluid Dynamics (CFD) for Nuclear Reactor Safety Applications - Workshop Proceedings, CFD4NRS-3 - Experimental Validation and Application of CFD and CMFD Codes to Nuclear Reactor Safety Issues
- Imprint Pagination
- 1231 p.
- Journal Page Range
- p. 100
- Report number
- NEA-CSNI-R--2011-14
Conference
- Title
- Computational Fluid Dynamics (CFD) for Nuclear Reactor Safety Applications - Experimental Validation and Application of CFD and CMFD Codes to Nuclear Reactor Safety Issues
- Acronym
- CFD4NRS-3
- Dates
- 14-16 Sep 2010
- Place
- Bethesda, Maryland (United States); Washington, DC (United States)
INIS
- Country of Publication
- Nuclear Energy Agency of the OECD (NEA)
- Country of Input or Organization
- Nuclear Energy Agency of the OECD (NEA)
- INIS RN
- 44089357
- Subject category
- S42: ENGINEERING; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- ACCURACY; C CODES; COMPARATIVE EVALUATIONS; COMPUTERIZED SIMULATION; HEAT TRANSFER; MATHEMATICAL MODELS; MESH GENERATION; NUCLEAR POWER PLANTS; PWR TYPE REACTORS; REACTOR SAFETY; TEMPERATURE DEPENDENCE; TEMPERATURE DISTRIBUTION; THERMAL HYDRAULICS; THERMAL SHOCK; THERMOCOUPLES; THREE-DIMENSIONAL CALCULATIONS; TRANSIENTS; TURBULENCE; TURBULENT FLOW; VALIDATION
- Descriptors DEC
- COMPUTER CODES; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; EVALUATION; FLUID FLOW; FLUID MECHANICS; HYDRAULICS; MEASURING INSTRUMENTS; MECHANICS; NUCLEAR FACILITIES; POWER PLANTS; POWER REACTORS; REACTORS; SAFETY; SIMULATION; TESTING; THERMAL POWER PLANTS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 2 refs.