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Title: Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

Abstract

A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle canmore » be achieved with the lower enrichment fuel.« less

Authors:
Publication Date:
Research Org.:
Idaho National Laboratory (INL)
Sponsoring Org.:
DOE - NE
OSTI Identifier:
912445
Report Number(s):
INL/CON-07-12090
TRN: US0800397
DOE Contract Number:  
DE-AC07-99ID-13727
Resource Type:
Conference
Resource Relation:
Conference: 2007 International Congress on Advances in Nuclear Power Plants,Nice, France,05/13/2007,05/18/2007
Country of Publication:
United States
Language:
English
Subject:
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; CARBIDES; CERAMICS; DESIGN; FABRICATION; FUEL RODS; IRRADIATION; ISOTOPE SEPARATION; NEUTRONS; NUCLEAR POWER PLANTS; REACTOR CORES; SAFETY MARGINS; SILICON CARBIDES; TEMPERATURE COEFFICIENT; TESTING; THERMAL NEUTRONS; TRANSIENTS; URANIUM; WATER; Very High Temperature Reactor, fuel design

Citation Formats

Sterbentz, James W. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core. United States: N. p., 2007. Web.
Sterbentz, James W. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core. United States.
Sterbentz, James W. Tue . "Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core". United States. doi:. https://www.osti.gov/servlets/purl/912445.
@article{osti_912445,
title = {Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core},
author = {Sterbentz, James W},
abstractNote = {A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue May 01 00:00:00 EDT 2007},
month = {Tue May 01 00:00:00 EDT 2007}
}

Conference:
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