Dynamic Analysis of the Next Generation Molten-Salt Breeder Reactor System

Journal Article · Sat Jul 01 00:00:00 EDT 2017 · Transactions of the American Nuclear Society

OSTI ID:23050350

Chvala, O.; Lish, M. R.; Singh, V.; Upadhyaya, B. R. ^[1]

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1. University of Tennessee, Department of Nuclear Engineering, Knoxville, TN 37996-2300 (United States)

The conceptual design of a 1000 MWe two-fluid molten-salt breeder reactor (MSBR) station was performed during 1966-1970 at Oak Ridge National Laboratory (ORNL) with the goal of using the U-233 fuel cycle. This modular design with four reactor modules has an electrical output of 250 MW/module. The abundance of Th-232 was attractive in converting it to U-233 in a thermal breeder reactor. The two-fluid reactor is graphite-moderated 'with a {sup 7}LiF-BeF{sub 2}-UF{sub 4} fuel salt circulated through the core and a {sup 7}LiF-ThF{sub 4}-BeF{sub 2} blanket salt circulated through separate flow channels distributed throughout the core, as well as in a surrounding under-moderated region'. This reactor design has the advantages of not producing Pu isotopes and easy chemical separation of U-233 in the blanket material. The low fuel cost and the graphite replacement cost are attractive for such a design, with an estimated cost of $141/kW at 1968-dollar value. Potential non-proliferation features of MSBR, the low costs of raw materials, and established chemical separation of fission products are important factors in further development of MSBRs as an alternative to current light water reactors. These reactors are often referred to as liquid fluoride thorium reactor (LFTR). The current research on the modeling of the MSBR is based on an earlier work on the preliminary dynamic model of the MSBR concept. Revisions are made to the original model. The physics parameters have been modified for the current modeling effort based on additional information including more recent values of model parameters from and several typographical misprints in equations and values. Details of this effort will be presented in a full follow-up paper. This work presents the detailed nodal models of reactor fluid core, blanket fluid loop, graphite moderator, fuel fluid heat exchanger, and the blanket fluid heat exchanger. A set of simulations has been performed to validate the model using the commercial software MATLAB{sup TM}-Simulink. Simulations were performed for the cases of source perturbation and reactivity perturbation. The model is inherently nonlinear. A description of the MSBR conceptual design, along with a tabulation of relevant system parameters, is presented. The results of simulation demonstrate stable behavior of the MSBR dynamics, temperature feedback effects, and load-following capability of the MSBR system.

• OSTI ID: 23050350
• Journal Information: Transactions of the American Nuclear Society, Vol. 116; Conference: 2017 Annual Meeting of the American Nuclear Society, San Francisco, CA (United States), 11-15 Jun 2017; Other Information: Country of input: France; 4 refs.; available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US); ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
BERYLLIUM FLUORIDES
BREEDER REACTORS
COMPUTERIZED SIMULATION
FISSION PRODUCTS
GRAPHITE
HEAT EXCHANGERS
LIQUIDS
LITHIUM FLUORIDES
MOLTEN SALTS
ORNL
PERTURBATION THEORY
PLUTONIUM ISOTOPES
RAW MATERIALS
REACTOR DESIGN
THORIUM 232
THORIUM FLUORIDES
URANIUM 233
URANIUM TETRAFLUORIDE
WATER COOLED REACTORS
WATER MODERATED REACTORS