Molten salt reactor

Example of a molten salt reactor scheme

A molten salt reactor (MSR) is a class of generation IV nuclear fission reactor in which the primary nuclear reactor coolant, or even the fuel itself, is a molten salt mixture. MSRs can run at higher temperatures than water-cooled reactors for a higher thermodynamic efficiency, while staying at low vapour pressure. [1]

The nuclear fuel may be solid or dissolved in the coolant. In many designs the nuclear fuel dissolved in the coolant is uranium tetrafluoride (UF4). The fluid becomes critical in a graphite core that serves as the moderator. Some solid-fuel designs propose ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high temperature cooling. The salts are much more efficient than compressed helium (another potential coolant in Generation IV reactor designs) at removing heat from the core, reducing the need for pumping and piping and reducing the core size.

The concept was established in the 1950s. The early Aircraft Reactor Experiment was primarily motivated by the small size that the design could provide, while the Molten-Salt Reactor Experiment was a prototype for a thorium fuel cycle breeder nuclear power plant. The increased research into Generation IV reactor designs included a renewed interest in the technology. [2]


Aircraft reactor experiment

Aircraft Reactor Experiment building at ORNL. It was later retrofitted for the MSRE.

Extensive research into molten salt reactors started with the U.S. aircraft reactor experiment (ARE) in support of the U.S. Aircraft Nuclear Propulsion program. The ARE was a 2.5 MWth nuclear reactor experiment designed to attain a high energy density for use as an engine in a nuclear-powered bomber.

The project included experiments, including high temperature reactor and engine tests collectively called the Heat Transfer Reactor Experiments: HTRE-1, HTRE-2 and HTRE-3 at the National Reactor Test Station (now Idaho National Laboratory) as well as an experimental high-temperature molten salt reactor at Oak Ridge National Laboratory – the ARE.

The ARE used molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, moderated by beryllium oxide (BeO). Liquid sodium was a secondary coolant.

The experiment had a peak temperature of 860 °C. It produced 100 MWh over nine days in 1954. This experiment used Inconel 600 alloy for the metal structure and piping. [3]

After ARE, another reactor was operated at the Critical Experiments Facility of the Oak Ridge National Laboratory in 1957. It was part of the circulating-fuel reactor program of the Pratt & Whitney Aircraft Company (PWAC). This was called the PWAR-1, the Pratt and Whitney Aircraft Reactor-1. The experiment was run for only a few weeks and at essentially zero nuclear power, but it reached criticality. The operating temperature was held constant at approximately 675 °C (1,250 °F). The PWAR-1 used NaF-ZrF4-UF4 as the primary fuel and coolant, making it one of the three critical molten salt reactors ever built. [4]

Molten-salt reactor experiment

MSRE plant diagram

Oak Ridge National Laboratory (ORNL) took the lead in researching the MSR through the 1960s. Much of their work culminated with the Molten-Salt Reactor Experiment (MSRE). The MSRE was a 7.4 MWth test reactor simulating the neutronic "kernel" of a type of epithermal thorium molten salt breeder reactor called the liquid fluoride thorium reactor. The large (expensive) breeding blanket of thorium salt was omitted in favor of neutron measurements.

The MSRE was located at ORNL. Its piping, core vat and structural components were made from Hastelloy-N, moderated by pyrolytic graphite. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-29-5-1). The graphite core moderated it. Its secondary coolant was FLiBe (2LiF-BeF2). It reached temperatures as high as 650 °C and operated for the equivalent of about 1.5 years of full power operation.

Oak Ridge National Laboratory molten salt breeder reactor

The culmination of the Oak Ridge National Laboratory research during the 1970–1976 timeframe resulted in a proposed molten salt breeder reactor (MSBR) design which would use LiF-BeF2-ThF4-UF4 (72-16-12-0.4) as fuel. It was to be moderated by graphite with a 4-year replacement schedule. The secondary coolant was to be NaF-NaBF4. Its peak operating temperature was to be 705 °C. [5] Despite the success, the MSR program closed down in the early 1970s in favor of the liquid metal fast-breeder reactor ( LMFBR), [6] after which research stagnated in the United States. [7] [8] As of 2011, the ARE and the MSRE remained the only molten-salt reactors ever operated.

The MSBR project received funding until 1976. Inflation-adjusted to 1991 dollars, the project received $38.9 million from 1968 to 1976. [9]

Officially, the program was cancelled because:

  • The political and technical support for the program in the United States was too thin geographically. Within the United States, only in Oak Ridge, Tennessee, was the technology well understood. [6]
  • The MSR program was in competition with the fast breeder program at the time, which got an early start and had copious government development funds allocated to many parts of the United States. When the MSR development program had progressed far enough to justify an expanded program leading to commercial development, the AEC could not justify the diversion of substantial funds from the LMFBR to a competing program. [6]

Oak Ridge National Laboratory denatured molten salt reactor (DMSR)

In 1980, the engineering technology division at Oak Ridge National Laboratory published a paper entitled "Conceptual Design Characteristics of a Denatured Molten-Salt Reactor with Once-Through Fueling." In it, the authors "examine the conceptual feasibility of a molten-salt power reactor fueled with denatured uranium-235 (i.e. with low-enriched uranium) and operated with a minimum of chemical processing." The main priority behind the design characteristics is proliferation resistance. [10] Lessons learned from past projects and research at ORNL were considered. Although the DMSR can theoretically be fueled partially by thorium or plutonium, fueling solely with low enriched uranium (LEU) helps maximize proliferation resistance.

Another important goal of the DMSR was to minimize R&D and to maximize feasibility. The Generation IV international Forum (GIF) includes "salt processing" as a technology gap for molten salt reactors. [11] The DMSR requires minimal chemical processing because it is a burner rather than a breeder. Both reactors built at ORNL were burner designs. In addition, the choices to use graphite for neutron moderation and enhanced Hastelloy-N for piping simplify the design and reduce R&D.

United Kingdom

The UK's Atomic Energy Research Establishment (AERE) were developing an alternative MSR design across its National Laboratories at Harwell, Culham, Risley and Winfrith. AERE opted to focus on a lead-cooled 2.5 GWe Molten Salt Fast Reactor (MSFR) concept using a chloride. [12] They also researched the option of helium gas as an alternative coolant. [13] [14]

The UK MSFR would be fuelled by plutonium, a fuel considered to be 'free' by the program's research scientists, because of the UK's plutonium stockpile.

Despite their different designs, ORNL and AERE maintained contact during this period with information exchange and expert visits. Theoretical work on the concept was conducted between 1964 and 1966, while experimental work was ongoing between 1968 and 1973. The program received annual government funding of around £100,000-£200,000 (equivalent to £2m-£3m in 2005). This funding came to an end in 1974, partly due to the success of the Prototype Fast Reactor at Dounreay which was considered a priority for funding as it went critical in the same year. [12]

AERE reports and findings from its MSR Program conducted in the 1960s and 1970s are available for public viewing at the UK National Archives in Kew, London. [12]

Soviet Union

In the USSR, a molten-salt reactor research program was started in the second half of the 1970s at the Kurchatov Institute. It included theoretical and experimental studies, particularly the investigation of mechanical, corrosion and radiation properties of the molten salt container materials. The main findings supported the conclusion that there were no physical nor technological obstacles to the practical implementation of MSRs. [15] A reduction in activity occurred after 1986 due to the Chernobyl accident, along with a general stagnation of nuclear power and the nuclear industry. [16](p381)

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