Rod bundle thermal-hydraulic and melt progression analysis of CORA severe fuel damage experiments

Journal Article · Fri Apr 01 04:00:00 EST 1994 · Nuclear Science and Engineering; (United States)

OSTI ID:7110188

Suh, K Y ^[1]

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1. Fauske Associates, Burr Ridge, IL (United States)

An integral, fast-running computational model is developed to simulate the thermal-hydraulic and melt progression behavior in a nuclear reactor rod bundle under severe fuel damage conditions. This consists of the submodels for calculating steaming from the core, hydrogen formation, heat transfer in and out of the core, cooling from core spray or injection, and, most importantly, fuel melting, relocation, and freezing with chemical interactions taking place among the material constituents in a degrading core. The integral model is applied to three German severe fuel damage tests to analyze the core thermal and melt behavior: CORA-16 (18-rod bundle and slow cooling), CORA-17 (18-rod bundle and quenching), and CORA-18 (48-rod bundle and slow cooling). Results of the temperature response of the fuel rods, the channel box, and the absorber blade; hydrogen generation from the fuel rod and the channel box; and core material eutectic formation, melt relocation, and blockage formation are discussed. Reasonable agreement is observed for component temperatures at midelevation where prediction and measurement uncertainties are minimal. However, discrepancies or uncertainties are noticed for hydrogen generation and core-melt progression. The experimentally observed peak generation of hydrogen upon reflooding is not able to be reproduced, and the total amount generated is generally underpredicted primarily because of the early relocation of the Zircaloy fuel channel box and cladding. Also, difficulties are encountered in the process of assessing the core-melt formation and the relocation model because of either modeling uncertainties or a lack of definitive metallurgical data as a function of time throughout the transient.

• OSTI ID: 7110188
• Journal Information: Nuclear Science and Engineering; (United States), Journal Name: Nuclear Science and Engineering; (United States) Vol. 116:4; ISSN NSENAO; ISSN 0029-5639
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTERIZED SIMULATION
ELEMENTS
ENERGY TRANSFER
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
HYDROGEN
MATHEMATICAL MODELS
MECHANICS
MELTDOWN
MITIGATION
NONMETALS
REACTOR ACCIDENTS
ROD BUNDLES
SIMULATION