Thermal analysis of uranium zirconium hydride fuel using a lead-bismuth gap at LWR operating temperatures

Next generation nuclear technology calls for more advanced fuels to maximize the effectiveness of new designs. A fuel currently being studied for use in advanced light water reactors (LWRs) is uranium zirconium hydride (UZH), a fuel currently being used in the popular TRIGA research reactors. UZH is being considered because unlike the current fuel of choice, uranium dioxide, it is metal based and therefore better able to transfer the heat out of the fuel that is coming from fission. This can lead to lower operating temperatures which will reduce the amount of fission gas release to negligible quantities, eliminate cracking, and reduce the internal energy of the fuel. Furthermore, it is hoped the UZH will be better able to attain higher burnups, partly because of the presence of the strong moderator hydrogen, and thus will help better utilize resources and reduce the volume of nuclear waste produced. In order for UZH to be viable as a fuel it is recommended that the peak central temperature of the fuel be maintained below 650°C, at which point swelling due to void formation around the uranium atoms becomes a concern. In order to keep temperature below this level it has been proposed that lead-bismuth eutectic (LBE) be used as the gap material instead of helium. In order to ensure that the properties of UZH while using a LBE gap, specifically the thermal conductivity, do not degrade to the point of the fuel not being viable, an experiment was designed and put into the MIT research reactor. The initial results show a decreasing trend in thermal conductivity, albeit with much of this change considered to be because of the many thermal cycles the experiment underwent while in the reactor.

Department dc:contributor.department

Massachusetts Institute of Technology. Dept. of Nuclear Science and Engineering.

Grantor dc:publisher

Massachusetts Institute of Technology

Year dc:date.issued

2012

Author dc:creator

• Ensor, Brendan M. (Brendan Melvin)

Advisor dc:contributor.advisor

• Gordon Kohse.

• Nuclear Science and Engineering.

Statement dc:rights

• M.I.T. theses are protected by copyright. They may be viewed from this source for any purpose, but reproduction or distribution in any format is prohibited without written permission. See provided URL for inquiries about permission.

Licence dc:rights.uri

• http://dspace.mit.edu/handle/1721.1/7582

Language dc:language.iso

eng

Handle dc:identifier.uri

http://hdl.handle.net/1721.1/76937

OAI identifier oai:identifier

oai:dspace.mit.edu:1721.1/76937