Doctoral Dissertations

Keywords and Phrases

LFTR; Liquid fluoride thorium reactor; Molten salt reactor; MSR; Nuclear reactor physics; Thorium fuel cycle

Abstract

"Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIGEN2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/232Th) = 2.01 % was the candidate fuel for this reactor core"--Abstract, page iii.

Advisor(s)

Alajo, Ayodeji Babatunde

Committee Member(s)

Lee, Hyoung-Koo
Castano, Carlos H.
Liu, Xin (Mining & Nuclear Engr)
Medvedeva, Julia E.

Department(s)

Mining and Nuclear Engineering

Degree Name

Ph. D. in Nuclear Engineering

Publisher

Missouri University of Science and Technology

Publication Date

Fall 2015

Pagination

xiii, 101 pages

Note about bibliography

Includes bibliographic references (pages 71-74).

Rights

© 2015 Safwan Qasim Mohammad Jaradat, All rights reserved.

Document Type

Dissertation - Open Access

File Type

text

Language

English

Library of Congress Subject Headings

Molten salt reactors
Nuclear fuels
Thorium
Neutron flux -- Measurement

Thesis Number

T 10825

Electronic OCLC #

936207257

Comments

Thesis does not contain page x.

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