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Advanced fuels modeling: Evaluating ...

Hallman, Luther, Jr.

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Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4.

紀錄類型:	書目-語言資料,印刷品 : Monograph/item
正題名/作者:	Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4./
作者:	Hallman, Luther, Jr.
面頁冊數:	96 p.
附註:	Source: Masters Abstracts International, Volume: 52-02.
Contained By:	Masters Abstracts International52-02(E).
標題:	Engineering, Nuclear. -
電子資源:	http://pqdd.sinica.edu.tw/twdaoapp/servlet/advanced?query=1544594
ISBN:	9781303353345

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4.
Hallman, Luther, Jr.

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4. - 96 p.

Source: Masters Abstracts International, Volume: 52-02.

Thesis (M.S.)--University of South Carolina, 2013.

Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism.

ISBN: 9781303353345Subjects--Topical Terms:

1043651
Engineering, Nuclear.

Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4.
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245 1 0 $a Advanced fuels modeling: Evaluating the steady-state performance of carbide fuel in helium-cooled reactors using FRAPCON 3.4.
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500 $a Adviser: Travis W. Knight.
502 $a Thesis (M.S.)--University of South Carolina, 2013.
520 $a Uranium carbide (UC) has long been considered a potential alternative to uranium dioxide (UO2) fuel, especially in the context of Gen IV gas-cooled reactors. It has shown promise because of its high uranium density, good irradiation stability, and especially high thermal conductivity. Despite its many benefits, UC is known to swell at a rate twice that of UO2. However, the swelling phenomenon is not well understood, and we are limited to a weak empirical understanding of the swelling mechanism.
520 $a One suggested cladding for UC is silicon carbide (SiC), a ceramic that demonstrates a number of desirable properties. Among them are an increased corrosion resistance, high mechanical strength, and irradiation stability. However, with increased temperatures, SiC exhibits an extremely brittle nature. The brittle behavior of SiC is not fully understood and thus it is unknown how SiC would respond to the added stress of a swelling UC fuel.
520 $a To better understand the interaction between these advanced materials, each has been implemented into FRAPCON, the preferred fuel performance code of the Nuclear Regulatory Commission (NRC); additionally, the material properties for a helium coolant have been incorporated. The implementation of UC within FRAPCON required the development of material models that described not only the thermophysical properties of UC, such as thermal conductivity and thermal expansion, but also models for the swelling, densification, and fission gas release associated with the fuel's irradiation behavior.
520 $a This research is intended to supplement ongoing analysis of the performance and behavior of uranium carbide and silicon carbide in a helium-cooled reactor.
590 $a School code: 0202.
650 4 $a Engineering, Nuclear. $3 1043651
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710 2 $a University of South Carolina. $b Nuclear Engineering. $3 2099533
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792 $a 2013
793 $a English
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