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(Reference retrieved automatically from Web of Science through information on FAPESP grant and its corresponding number as mentioned in the publication by the authors.)

Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel

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Author(s):
Lopes, Denise A. [1] ; Zimmermann, Angelo J. O. [2] ; Silva, Selma L. [2] ; Piqueira, J. R. C. [1]
Total Authors: 4
Affiliation:
[1] Univ Sao Paulo, Escola Politecn, Ave Prof Mello Moraes 2463, BR-05508030 Sao Paulo, SP - Brazil
[2] CEA, CTMSP, Estr Sorocaba Ipero, Km 12, 5, BR-18560000 Ipero, SP - Brazil
Total Affiliations: 2
Document type: Journal article
Source: JOURNAL OF NUCLEAR MATERIALS; v. 473, p. 136-142, MAY 2016.
Web of Science Citations: 3
Abstract

Uranium alloys in a monolithic form have been considered attractive candidates for high density nuclear fuel. However, this high-density fissile material configuration keeps the volume permitted for the retention of fission products at a minimum. Additionally, the monolithic nuclear fuel has a peculiar configuration, whereby the fuel is in direct contact with the cladding. How this fuel configuration will retain fission products and how this will affect its integrity under various physical conditions - such as thermal cycling - are some of the technological problems for this new fuel. In this paper, the effect of out-of- pile thermal cycling is studied for a monolithic fuel plate produced by a hot co-rolling method using U-10Mo (wt %) as the fuel alloy and Zircaloy-4 as the cladding material. After performing 10 thermal cycles from 25 to 400 degrees C at a rate of 1 degrees C/min (similar to 125 h), the fuel alloy presented several fractures that were observed to occur in the last three cycles. These cracks nucleated approximately in the center of the fuel alloy and crossed the interdiffusion zone initiating an internal crack in the cladding. The results suggest that the origin of these fractures is the thermal fatigue of the U-10Mo alloy caused due to the combination of two factors: (i) the high difference in the thermal expansion coefficient of the fuel and of the cladding material, and (ii) the bound condition of fuel/cladding materials in this fuel element configuration. (C) 2016 Elsevier B.V. All rights reserved. (AU)

FAPESP's process: 10/11113-2 - Critical heat flux experiment specifications, test section basic and detailed design, data acquisition analysis and correlations for a small Pressurized Water Reactor (PWR) (up to 50 MWt)
Grantee:José Roberto Castilho Piqueira
Support Opportunities: Research Projects - Thematic Grants