The purpose of this project is to design a Molten Salt Reactor (MSR) that utilizes resources much more efficiently than other nuclear reactors, the thermal output of which can be used for efficient and large-scale production of hydrogen.
Hydrogen is generally considered as a clean energy carrier for future transport concepts. It could be used for power generation in fuel cells, added to crude oil to produce clean and lighter gasoline, or used as an ingredient to synthetic fuel production. Sustainable mobility based on hydrogen can only be achieved if hydrogen could be produced at large scale using (virtually) inexhaustible energy resources like solar, wind, hydro and nuclear. If a nuclear reactor could produce heat at a temperature of 700-1000 °C then hydrogen can be produced using one of the following processes:
Thermo-chemical water splitting. The most promising concepts use the Sulfur- Iodine cycle or the Calcium-Bromium (UT-3) cycle. The first employs three chemical reactions at temperatures of 850, 120 and 450 °C, while the latter has four at 670, 560, 760 and 210 °C. Both cycles can reach high efficiency (> 50%).
High temperature electrolysis of water. In this process steam at high temperature (700-900 °C) is dissociated, which is close to the reverse reaction of power generation in a solid oxide fuel cell.
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.
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.
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 °C. 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.
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.
The MSR in combination with the thorium fuel cycle has many advantages:
Amongst others, there are some main topics that need special attention:
Results of the research can be found in the PhD thesis.
|For more information, please contact j.l.kloosterman (at) tudelft.nl|