Fuel resources

Although nuclear hydrogen production is a very attractive option, the nuclear reactors currently in operation as well as the leading candidate for hydrogen production (the Very High Temperature Reactor), utilize only 1% of the available natural uranium. Continuing this way, known resources cover less than 60 years of consumption and cannot support nuclear expansion in the next decades. This possible shortage of fissile uranium calls for nuclear reactors that convert either the non-fissile uranium isotope U-238 to the fissile plutonium isotope Pu-239, or the non-fissile thorium isotope Th-232 to the fissile U-233. Especially the latter option is attractive, because thorium is estimated to be four times more abundant than uranium in the earth's crust, and the usage of thorium considerably reduces the production of long-lived nuclear waste. The known thorium reserves could fuel the existing nuclear reactor park for several tens of thousand of years and the estimated additional resources might over double or triple this period.

The Molten Salt Reactor, which will be discussed in the next section, is the most attractive reactor for the usage of thorium to produce electricity and process heat that can be used for production of hydrogen.

The Molten Salt Reactor

General Description

A Molten Salt Reactor consists of a graphite core with vertical channels through which a salt-fuel mixture flows at ambient pressure. Only in the core, neutrons released in a fission event can moderate to thermal energies and initiate new fissions; outside the core no fission takes place. Because the salt is used to remove the heat from the core and to circulate the fuel, a fraction of it can be diverged to a chemical processing plant to extract fission products and add fresh fuel. This means that fewer neutrons are lost by parasitic neutron capture, which enables the MSR to convert non-fissile isotopes to fissile ones. In this way, otherwise useless uranium and thorium isotopes can be used to fuel the reactor. Via a secondary loop containing a coolant salt as well, the heat is transferred to a heat exchanger or steam generator, depending on the energy conversion system. The coolant salt can also be used to feed a hydrogen production plant.

Scheme of the Molten Salt Breeder Reactor (MSBR).

History

The Oak Ridge National Laboratory (ORNL) operated the Molten Salt Reactor Experiment (MSRE) from 1965 to 1970. It used UF₄ made of 32% enriched uranium mixed with LiF, BeF₂, and ZrF₄ to make up the salt in the primary circuit. The salt in the secondary circuit consisted of a LiF₂-BeF₂ mix named FLIBE. In 1968 the fuel composition changed to U-233, making the MSRE the first reactor in the world using this fuel. From this reactor it is known that fluoride salts do not penetrate graphite if the pores are made small enough. Furthermore, INOR-8 (Hastelloy-N), a specially developed nickel-molybdenum-iron-chromium alloy, proved very resistant to corrosion by fluoride salts at temperatures up to 800 ^oC. The MSRE provided experimental data that can be used for validating and benchmarking nuclear data and calculation code systems.

In the beginning of the 1970's, after shut-down of the MSRE, the Molten Salt Breeder Reactor (see the Figure) was designed. The objective was to develop the most efficient U-233 breeder in a thorium cycle. It uses a mixture of fluoride salts of beryllium, lithium, thorium and uranium in the primary circuit and NaBF₄-NaF in the secondary circuit. Because of budget constraints, the development of this reactor was stopped in favor of the Liquid Metal Fast Breeder Reactor (LMFBR), which had major federal funding at that time due to high expectations regarding breeding gain and plutonium production.

Current status

The MSR is one of the six reactor designs being studied in the framework of the Generation-IV program in strong competition with the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). In a fast reactor, fission events are initiated by fast neutrons, liberating more fission neutrons per event, which can be used to breed new fissile material or transmute nuclear waste. Molten Salt reactors, on the other hand, utilize a thermal neutron spectrum (although epithermal and fast reactors could be designed as well), in which breeding is established through the excellent neutron economy. Neutron absorption in graphite is negligible, while the absorbing fission products can be removed on-line from the fluid fuel in a chemical processing plant (see the Figure). This leaves just enough neutrons to breed new fissile material. In a thermal reactor, of all nuclides, U-233 releases most fission neutrons, making it the preferred fuel for the MSR. Nowadays, most research groups focus on the capability of the MSR as a nuclear waste incinerator, which is also the main objective of the Generation-IV program. In this project, we will focus on the MSR as a "self-breeder", producing its own fuel from cheap and abundant thorium without generating a surplus of proliferation sensitive fissile material, like all other reactors in the world do.

Advantages

The MSR in combination with the thorium fuel cycle has many advantages:

1. Fluoride inorganic salts are used as a carrier for the fuel and as a coolant. They are among the most stable of chemical compounds and have proven stable under reactor operating conditions. They have a high solubility for actinides, very low vapor pressure, and good heat transfer properties. Furthermore, they do not react with air or water, and are inert to some commonly used structural materials.

2. Soluble fission products can be removed on-line in a chemical processing plant, while non-soluble fission products and the noble metals can be extracted from the salt by helium bubbling. This enhances the neutron economy. Together with the large number of neutrons liberated in U-233 fission events, new fissile material can be bred from abundantly available thorium.

3. There are no mechanical valves in the salt circuit. Flow is blocked by plugs of frozen salt cooled by electrical fans. If the salt heats up to levels above design values or if the power supply fails, the plugs will melt and the salt will be drained into storage drums cooled by natural convection (see the Figure).

4. A fast excursion of the fuel temperature will lead to salt expansion providing instantaneous negative reactivity feedback, which will slow down or completely stop the fission process. Although heating of the graphite moderator will generally introduce positive reactivity, this process is much slower and can easily be controlled. Furthermore, a fuel salt temperature too high will always lead to drainage of the fuel into passively cooled storage tanks.

5. The primary and secondary circuits are operated under ambient pressure, which is considered a very important safety feature.

6. The thorium fuel cycle produces much less long-lived nuclear waste. Compared with the standard once-through fuel cycle in a Light Water Reactor (LWR), a thorium fueled MSR produces 4,000 times less neptunium, plutonium, americium and curium. Plutonium production is reduced even with a factor of 10,000.

7. Among all nuclear reactors, the MSR is most suited to utilize the thorium cycle. Neutron capture by Th-232 produces Pa-233, which decays with a half life of 27 days to U-233. To avoid Pa-233 capturing an extra neutron, which would produce the non-fissile U-234, part of it can easily be stored in a hold-up tank to let it decay to U-233. This enhances the breeding process, which makes the MSR, in combination with its excellent neutron economy, the most attractive reactor for using thorium.