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Massachusetts Institute of Technology

Thermomechanical analysis of innovative nuclear fuel pin designs

Abstract

dc:description.abstract

One way to increase the power of a nuclear reactor is to change the solid cylindrical fuel to Internally and Externally Cooled (I&EC) annular fuel, and adjust the flow and the core inlet coolant temperature. The switch to annular fuel allows for a 20% increase in core power density without changing the assembly size or the control rod placement. Such an approach is being considered for the Korean reactor OPR1400. The analysis of I&EC fuel with a modified version of FRAPCON-3.3 revealed that an uneven heat flux split between the outer and inner surfaces may develop which does not yield an optimal design. After optimization, it is found that excessive cladding oxidation on the inner channel of the hottest fuel pin may occur due to excessive internal heat flux, which is controlled by the gap conductance. As the gaps close asymmetrically, the MDNBR limit of 1.3 could also be violated. At the uprated power, control of the gap sizes is needed in order to satisfy the thermal-hydraulic requirements. One solution is to increase the flow rate to increase the MDNBR and to reduce the coolant temperature to decrease the cladding oxidation. Reduced-moderation Water Reactor (RMWR) is a boiling water reactor proposed to operate with mixed oxide fuel and, a harder neutron spectrum and higher local fuel burnup compared to the traditional Light Water Reactors. The fissile content of the fuel is concentrated in two pancake like regions in the core separated by a blanket (fertile-only) region.

Degree

thesis:*
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
2010

Author and committee

dc:creator, dc:contributor.*
Author dc:creator
  • Lerch Andrew (Andrew J.)
Advisor dc:contributor.advisor
  • Mujid S. Kazimi.

Subjects

dc:subject × 1

Rights

dc:rights
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.
Language dc:language.iso
eng

Identifiers

dc:identifier.*
Handle dc:identifier.uri
http://hdl.handle.net/1721.1/58460
OAI identifier oai:identifier
oai:dspace.mit.edu:1721.1/58460

Chain of custody

source
Harvested from
MIT
Base URL
dspace.mit.edu/oai/request
Last updated
2026-07-22
Source record
OAI-PMH GetRecord
citation

Lerch Andrew (Andrew J.). Thermomechanical analysis of innovative nuclear fuel pin designs. Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/58460