Design of a Molten Salt-cooled Nuclear Reactor for Sustainable Hydrogen Production

Objective

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.

Sustainable mobility

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:

  1. 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%).

  2. 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.

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.

MSBR
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 °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.

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.

Research Topics

Amongst others, there are some main topics that need special attention:

  1. The moving salt in the primary circuit contains the fuel as well. Calculation code systems designed for solid fuel reactors have to be adapted to take into account the fuel transport and the on-line fuel processing.
  2. Nuclear reactors can be controlled by the grace of delayed neutrons emitted by precursor atoms up to some tens of seconds after the fission event. Due to the circulating fuel, a fraction of these neutrons is lost, which effectively reduces the margin to prompt criticality (the point at which a rapid power excursion might take place). In a properly designed MSR, prompt feedback can be assured by the effect of fuel expansion. Therefore, knowledge about the density and temperature of the fuel salt and the temperature distribution in the core are important to predict accurately the dynamics of the reactor during normal operation and transients.
  3. The fuel breeding process in the MSR can be enhanced by designing a two-fluid reactor, having separate flows of driver fuel through the center of the core and breeder fuel (containing more thorium) through the blanket surrounding the driver zone. Because the driver fuel contains less thorium and more fissile uranium, the power feedback coefficients could benefit from this as well. This needs further research and optimization.

Description of the PhD Project

Work packages

  1. Literature study to old concepts, like the MSRE and the MSBR developed at the Oak Ridge National Laboratory in the 1960's and 1970's.
  2. Modification and development of computer codes to include the effects on the neutronics and thermal-hydraulics of the moving fuel, heat deposition in the graphite moderator, on-line fuel processing, etc. Benchmarking and validation using data of the MSRE experimental data.
  3. Static and dynamic design of a MSR using thorium. This includes the calculation of reactivity feedback coefficients and the transient behavior. Note that the total feedback is determined by the opposite effects of fuel expansion and graphite heating. Pump start-up and coast-down transients will be investigated as well as reactivity induced accidents (e.g. control rod withdrawal) and loss of flow and loss of heat sink. If necessary, measures to reduce the positive feedback due to graphite heating will be investigated, like adding a neutron absorber to the moderator or zoning the core in driver and breed regions.
  4. Fuel cycle studies on the MSR using thorium. In particular the start-up of the reactor with other fuel types (assuming no initial U-233 will be available) and the gradual shift towards the thorium fuel cycle will be investigated. The required fuel processing rates will be determined both for "self breeding" (producing just enough U-233 to operate the reactor) as well as for "surplus breeding" to start up other molten salt reactors.
  5. The optimal hydrogen production process will be selected and modeled. Coupling phenomena of the MSR to this hydrogen production plant will be investigated. The overall efficiency will be determined as well as the dynamic behavior of the whole system.


For more information, please contact  j.l.kloosterman@tudelft.nl Home