Purdue University Graduate School
Browse

NEUTRONIC ANALYSIS OF THORIUM-BASED MOLTEN SALT REACTOR

11

month(s)

26

day(s)

until file(s) become available

NEUTRONIC ANALYSIS OF THORIUM-BASED MOLTEN SALT REACTOR

thesis
posted on 2023-04-24, 14:49 authored by Seda Yilmaz KaygisizSeda Yilmaz Kaygisiz

Environmental concerns and the increase in energy demand with technology and innovations lead us to develop more efficient and environmentally friendly energy sources. With that purpose, Generation IV International Forum (GIF) Charter determined six advanced nuclear reactors. This thesis focuses on one of those reactors, Molten Salt Reactors (MSRs). MSR was chosen because of its outstanding feature; molten salt coolant/fuel. The fuel/coolant in this reactor type is in its molten salt form, which enables the reactor to reach a high temperature ( ̴ 600°C) with low pressure. As high temperature enhances thermal efficiency, low pressure makes the reactor safer. Besides, low pressure enables the reactor to be more economical because there is no need to use large pumps to maintain high pressure and no need for the pressure vessel, thereby decreasing the cost of construction and maintenance. Furthermore, molten salt reactors are inherently safer than light water reactors (LWRs) due to the molten salt fuel. Materials and structures are designed to tolerate the high-temperature molten salt, which means no risk of a core meltdown accident.

A comprehensive literature review has been made on molten salt reactors to have a broad knowledge of MSR design characteristics and the current developments in the reactor. The literature review highlighted the notable features of this reactor design; being inherently safer and economical and having high thermal efficiency. In addition, the literature review showed that there are many studies on MSRs with different designs and materials for different purposes. However, the current parametric studies on literature were mainly performed for single channels and limited materials, meaning there is limited knowledge of whole reactor core analysis. This observation led us to perform a complete MSR core analysis with various design parameters; core size, moderator and fuel/coolant materials, and core configurations (hexagonal and hexahedral lattice geometries). Considering the advantages of MSRs and the need for detailed work on this reactor type in literature, a parametric study on the reactor was performed under the thesis presented here. 

Monte Carlo N-Particle (MCNP) 6.2 code is chosen for the criticality and the flux simulations. The single-fluid double-zone thorium-based molten salt reactor (SD-TMSR) has been selected as a base model. Single fluid means molten salt is the same as for fuel and coolant. Various molten salt compositions have been investigated to observe the effect of different elements and isotopes on criticality. The active core has two zones with the same molten salt but different fuel/coolant channel diameters for each zone. The inner zone represents where the fission reaction occurs mainly, and the outer zone serves as a blanket that enables the breeding process of thorium. To determine the criticality behavior of the reactor with moderation, simulations were performed with various inner zone fuel channel radii, from 0.25 cm to 5 cm. In comparison with the channel radius of the inner one, the outer zone fuel channel radius is fixed at 5 cm. 

Additionally, graphite and BeO moderators were examined separately to decide the material for better moderation. On the other hand, the core configuration is essential to make a more compact reactor. Therefore, hexagonal and hexahedral lattice geometries were simulated with all other cases; different fuel channel diameters, molten salts, and moderators. Before the flux distribution and the heat transfer calculations, the best combinations of the parameters which reach the criticality with the less fuel inventory have been decided and used for further calculations. Finally, four fuel/coolant salts have been chosen for the thermal neutron flux distribution simulations. Results for the flux distribution were represented with 2D and 3D color graphs, and results for different salts were compared with 2D graphs for axial and radial directions. Besides, to obtain a general idea of the reactor's power density and thermal-hydraulics characteristics, heat transfer calculations were done for the hot channel as a transition from neutronics to thermal hydraulics for future studies. With those calculations, power density, an average mass flow rate, and core inlet/outlet temperatures were determined.

Funding

Turkish Republic of the National Ministry of Education

History

Degree Type

  • Master of Science

Department

  • Nuclear Engineering

Campus location

  • West Lafayette

Advisor/Supervisor/Committee Chair

Dr. Shripad T. Revankar, Chair

Additional Committee Member 2

Yunlin Xu

Additional Committee Member 3

David S. Koltick

Usage metrics

    Licence

    Exports

    RefWorks
    BibTeX
    Ref. manager
    Endnote
    DataCite
    NLM
    DC