NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

Conference · Tue Nov 01 04:00:00 EDT 2011

OSTI ID:1035896

& Herman Shen, Hakan Ozaltun

This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

• Research Organization: Idaho National Laboratory (INL)
• Sponsoring Organization: DOE - NA
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1035896
• Report Number(s): INL/CON-11-22390
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
ACCURACY
ALLOYS
CREEP
FABRICATION
FISSION
FUEL PLATES
IRRADIATION
MECHANICAL ENGINEERING
MELTING
REACTOR CORES
RERTR
RESEARCH AND TEST REACTORS
RESIDUAL STRESSES
STRESS RELAXATION
STRESSES
SWELLING
THERMAL CYCLING
THERMAL STRESSES
TRANSIENTS
URANIUM