RELAP5 model to simulate the thermal-hydraulic effects of grid spacers and cladding rupture during reflood

Technical Report · Fri Sep 01 00:00:00 EDT 1995

DOI:https://doi.org/10.2172/115100· OSTI ID:115100

Nithianandan, C. K.; Klingenfus, J. A.; Reilly, S. S. ^[1]

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1. B&W Nuclear Technologies, Lynchburg, VA (United States)

Droplet breakup at spacer grids and a cladding swelled and ruptured locations plays an important role in the cooling of nuclear fuel rods during the reflooding period of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). During the reflood phase, a spacer grid affects the thermal-hydraulic system behavior through increased turbulence, droplet breakup due to impact on grid straps, grid rewetting, and liquid holdup due to grid form losses. Recently, models to simulate spacer grid effects and blockage and rupture effects on system thermal hydraulics were added to the B&W Nuclear Technologies (BWNT) version of the RELAP5/MOD2 computer code. Several FLECHT-SEASET forced reflood tests, CCTF Tests C1-19 and C2-6, SCTF Test S3-15, and G2 Test 561 were simulated using RELAP5/MOD2-B&W to verify the applicability of the model at the cladding swelled and rupture locations. The results demonstrate the importance of modeling the thermal-hydraulic effects due to grids, and clad swelling and rupture to correctly predict the clad temperature response during the reflood phase of large break LOCA. The RELAP5 models and the test results are described in this paper.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Div. of Systems Technology; American Nuclear Society (ANS), La Grange Park, IL (United States); American Institute of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers (ASME), New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); European Nuclear Society (ENS), Bern (Switzerland); Atomic Energy Society of Japan, Tokyo (Japan); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 115100
• Report Number(s): NUREG/CP-0142-Vol.3; CONF-950904-Vol.3; ON: TI95017079; TRN: 95:023000
• Resource Relation: Conference: 7. international topical meeting on nuclear reactor thermal-hydraulics (Nureth-7), Saratoga Springs, NY (United States), 10-15 Sep 1995; Other Information: PBD: Sep 1995; Related Information: Is Part Of Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7, Volume 3, Sessions 12-16; Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)]; PB: 1001 p.
• Country of Publication: United States
• Language: English

Related Subjects

22 NUCLEAR REACTOR TECHNOLOGY
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING NOT INCLUDED IN OTHER CATEGORIES
99 MATHEMATICS
COMPUTERS
INFORMATION SCIENCE
MANAGEMENT
LAW
MISCELLANEOUS
PWR TYPE REACTORS
SAFETY ANALYSIS
LOSS OF COOLANT
COMPUTERIZED SIMULATION
FUEL ELEMENT FAILURE
MATHEMATICAL MODELS
HYDRAULICS
REACTOR SAFETY
NUMERICAL DATA