Koshi Mitachi

Three-Region Core Design for 200-MW(electric) Molten-Salt Reactor with Thorium-Uranium Fuel

In this paper, an improved design for a small molten-salt reactor (MSR) that uses neutron flux flattening, which is referred to as FUJI-U3, is proposed. This reactor is a 200-MW(electric) power reactor, and its core contains graphite (as the moderator) and fuel salt. The fuel salt is composed of ThF4 as the fertile material, 233UF4 as the fissile material, and LiF-BeF2 as both the solvent and heat transfer medium. A basic improvement in FUJI-U3 is the introduction of the design concept of a three-region core in order to avoid the replacement of graphite, which is achieved by reducing the maximum neutron flux. Since there is a limit for irradiation growth in graphite, this reduction in the maximum neutron flux contributes to a longer lifetime of the graphite. Based on calculations using the nuclear analysis code SRAC95 and the burnup analysis code ORIGEN2, it is concluded that there is no need to replace the graphite moderator of FUJI-U3 for 30 yr. Further, the chemical-processing interval of the fuel salt is studied for 7.5, 15, and 30 yr. An increase in this time interval will also contribute to reduce maintenance and cost.

Steady State Analysis of Molten Salt Reactor in Consideration of the Effect of Fuel Salt Flow

The Molten Salt Reactor (MSR) is a thermal neutron reactor with graphite moderation and operates on the thorium-uranium fuel cycle. The feature of the MSR is that fuel salt flows the inside of the reactor accompanying nuclear fission reaction. In the previous study, the authors had developed numerical model to simulate the effects of the fuel salt flow on the reactor characteristics. This paper applies the model to the steady state analysis of the small MSR system and estimates the effects of the fuel flow. The model consists of two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The following results are obtained: (1) the fuel salt flow affects the distributions of the delayed neutron precursors, especially long-lived one, and (2) the extension of residence time in the external loop system and the rise of fuel inflow temperature slightly show negative reactivity effects, decreasing neutron multiplication factor of the small MSR system.Copyright

Reactor Controllability of 3-Region-Core Molten Salt Reactor System: A Study on Load Following Capability

The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of ssile fuels (breeding). Thorium (Th) and uranium-233 (233 U) are fertile and ssile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development [1]. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the ssion reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. [2] and Shimazu et al. [3] developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow.Copyright

Temperature Distribution in Graphite Moderators and Fuel Salt in Core of Small Molten-Salt Power Reactor.

In the design of a small molten salt power reactor, one of the important problems is the heat deposition in graphite moderators, which are irradiated with high energy γ-ray and neutron flux. To remove the deposited heat adequately, fuel salt should be distributed appropriately in the reactor core.A numerical study was carried out in this paper, and the main concern was about the temperature distributions in the graphite moderator elements and the fuel salt. The velocity distribution of fuel salt in the reactor core was increased in proportion to the heat generation distribution which was given from the preceding neutronic design study. The heat conduction equation with appropriate boundary conditions was solved by finite element method to simulate heat transmission process in the graphite moderators. It was revealed that the maximum temperature of about 1, 150K appeared in the graphite element in the blanket. The maximum temperature gradient of about 4K/mm appeared in the element in the inner core (core I). It was also estimated that the velocity of fuel salt was about 1.0m/s in the core I, and 0.1m/s in the blanket. The thermal design study represented in this paper suggested that the heat deposition would not cause severe restrictions to construct and to run the small molten-salt power reactor.