Molten salt reactor: Wikis

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Molten salt reactor scheme.

A molten salt reactor (MSR) is a type of nuclear reactor where the primary coolant is a molten salt mixture, which can run high temperatures (for higher thermodynamic efficiency) while staying at low vapor pressure for reduced mechanical stress and increased safety, and is less reactive than molten sodium coolant. The nuclear fuel may be solid fuel rods, or dissolved in the coolant itself, which eliminates fuel fabrication, simplifies reactor structure, equalizes burnup, and allows online reprocessing.

In many designs the nuclear fuel is dissolved in the molten fluoride salt coolant as uranium tetrafluoride (UF4). The fluid becomes critical in a graphite core which serves as the moderator. Fluid fuel reactors have significantly different safety issues; the potential for major reactor accidents is reduced, while the potential for processing accidents is increased.[1]

More recent research has focused on the practical advantages of the high-temperature low-pressure primary cooling loop. Many modern designs rely on ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high temperature cooling. The salts are much more efficient at removing heat from the core, reducing the need for pumping, piping, and reducing the size of the core as these components are reduced in size.

The early Aircraft Reactor Experiment (1954) was primarily motivated by the small size that the design could provide, while the Molten-Salt Reactor Experiment (1965-69) was a prototype for a thorium fuel cycle breeder reactor nuclear power plant. One of the Generation IV reactor designs is a molten salt-cooled, solid-fuel reactor; the initial reference design is 1000 MWe with a deployment target date of 2025.

Another advantage of a small core is that it has fewer materials to absorb neutrons. The improved neutron economy makes neutrons available so that the Thorium 232 can breed into Uranium 233. So, its compact core makes the molten salt design particularly suitable for the thorium fuel cycle.

Contents

History

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The 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 US Aircraft Reactor Experiment (ARE). The US Aircraft Reactor Experiment was a 2.5 MWth nuclear reactor experiment designed to attain a high power density for use as an engine in a nuclear powered bomber. The project resulted in several experiments, three of which resulted in engine tests collectively called the Heat Transfer Reactor Experiments: HTRE-1, HTRE-2, and HTRE-3. One experiment used the molten fluoride salt NaF-ZrF4-UF4 (53-41-6 mol%) as fuel, was moderated by beryllium oxide (BeO), used liquid sodium as a secondary coolant, and had a peak temperature of 860 °C. It operated for a 1000 hour cycle in 1954. This experiment used Inconel 600 alloy for the metal structure and piping.

The Molten-Salt Reactor Experiment

MSRE plant diagram

Oak Ridge National Laboratory took the lead in researching the MSR through 1960s, and 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 an inherently safe epithermal thorium breeder reactor. It tested molten salt fuels of uranium and plutonium. The tested 233UF4 fluid fuel has a unique decay path that minimizes waste, with waste isotopes having half-lives under 50 years. The red-hot 650 °C temperature of the reactor could power high-efficiency heat engines such as gas turbines. 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 and its moderator was pyrolytic graphite. It went critical in 1965 and ran for four years. The fuel for the MSRE was LiF-BeF2-ZrF4-UF4 (65-30-5-0.1), the graphite core moderated it, and 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 reactor

The culmination of the Oak Ridge National Laboratory research during the 1970-76 timeframe resulted in an MSR design which would use LiF-BeF2-ThF4-UF4 (72-16-12-0.4) as fuel, was to be moderated by graphite with a 4 year replacement schedule, use NaF-NaBF4 as the secondary coolant, and have a peak operating temperature of 705 °C. [2]

The molten salt reactor offers many potential advantages: [3]

  • inherently safe design (safety by passive components and the strong negative temperature coefficient of reactivity)
  • using an abundant supply of Thorium to breed U-233 fuel.
  • much cleaner: the amount of waste fission products is 10 times less (per GWh) and requires containment for 100 times shorter time (300 years vs. tens of thousands of years)
  • can "burn" some problematic radioactive waste (with transuranic elements from traditional solid-fuel nuclear reactors)
  • possible even in small, even 2-8MW(th) or 1-3MW(electric). Submarine or aircraft size is possible
  • can react to load changes in less than 60 seconds (unlike "traditional" solid-fuel nuclear power plants)

Liquid salt very high temperature reactor

Research is currently picking up again for reactors that utilize molten salts for coolant. Both the traditional molten salt reactor and the Very High Temperature Reactor (VHTR) have been picked as potential designs to be studied under the Generation Four Initiative (GEN-IV). A version of the VHTR currently being studied is the Liquid Salt Very High Temperature Reactor (LS-VHTR),also commonly called the Advanced High Temperature Reactor (AHTR). It is essentially a standard VHTR design that uses liquid salt as a coolant in the primary loop, rather than a single helium loop. It relies on "TRISO" fuel dispersed in graphite. Early AHTR research focused on graphite would be in the form of graphite rods that would be inserted in hexagonal moderating graphite blocks, but current studies focus primarily on pebble-type fuel. The LS-VHTR has many attractive features, including: the ability to work at very high temperatures (the boiling point of most molten salts being considered are >1400 °C), low pressure cooling that can be used to more easily match hydrogen production facility conditions (most thermo chemical cycles require temperatures in excess of 750 °C), better electric conversion efficiency than a helium cooled VHTR operating at similar conditions, passive safety systems, and better retention of fission products in the event of an accident.

The Fuji MSR

The FUJI mini-MSR is a 100 MWe molten-salt-fueled Thorium fuel cycle thermal breeder reactor design, using technology similar to the Oak Ridge National Laboratory Reactor. It is being developed by a consortium including members from Japan, the U.S. and Russia. As a breeder reactor, it converts Thorium into nuclear fuels.[4] As a thermal-spectrum reactor, its neutron regulation is inherently safe. Like all molten salt reactors, its core is chemically inert, under low pressures to prevent explosions and toxic releases.[5] It would likely take 20 years to develop a full size reactor [6] but the project seems to lack funding.[7]

Technological issues

Molten-salt fueled reactors

The classic MSFR has been very exciting to many nuclear engineers. Its most prominent champion was Alvin Weinberg, who patented the light-water reactor and was a director of the U.S.'s Oak Ridge National Laboratory, a prominent nuclear research center.

Two concepts were investigated. The "two fluid" reactor had a high-neutron-density core that burned uranium-233 from the thorium fuel cycle. A blanket of thorium salt absorbed the neutrons and was eventually transmuted to 233U fuel. The weakness of the two-fluid design at the time of development was that its known designs included complex plumbing, and no suitable material was known to make the pipes. Ordinary steels and nickel alloys either absorbed too many neutrons or corroded too easily. Graphite was thought to be too brittle, and swells slightly under intense neutron exposure. Zirconium is sufficiently transparent to neutrons, but corrodes too easily when exposed to hot fluoride salts.

Two problems were subsequently solved by researchers at the Oak Ridge National Laboratory. The corrosion of the pipes was stopped by the addition of a trace amount of titanium to Hastelloy-N alloy.

The engineers discovered that by carefully sculpting the moderator rods (to get neutron densities similar to a core and blanket), and modifying the fuel reprocessing chemistry, both thorium and uranium salts could coexist in a simpler, less expensive yet efficient "one fluid" reactor.

The power reactor design produced by Weinberg's research group was similar to the MSRE above, which was designed to validate the risky hot, high-neutron-density "kernel" part of the "kernel and blanket" thorium breeder.

Technological advantages

The advantages cited by Weinberg and his associates at Oak Ridge National Laboratory include:

  • It's safe to operate and maintain: Molten fluoride salts are mechanically and chemically stable at sea-level pressures at intense heats and radioactivity. Fluoride combines ionically with almost any transmutation product, keeping it out of circulation. Even radioactive noble gases — notably xenon-135, an important neutron absorber — come out in a predictable, containable place, where the fuel is coolest and most dispersed, namely the pump bowl. Even given an accident, dispersion into a biome is unlikely. The salts do not burn in air or water, and the fluoride salts of the actinides and radioactive fission products are generally not soluble in water.
  • There's no high pressure steam in the core, just low-pressure molten salt. This means that the MSR's core cannot have a steam explosion, and does not need the most expensive item in a light water reactor, a high-pressure steam vessel for the core. Instead, there is a vat and low-pressure pipes (for molten salt) constructed of thick sheet metal. The metal is an exotic nickel alloy that resists heat and corrosion, Hastelloy-N, but there is much less of it, and the thin metal is less expensive to form and weld.
  • The thorium breeder reactor uses low-energy thermal neutrons, similarly to light water reactors. It is therefore much safer than the touchy fast-neutron breeder reactors that the uranium-to-plutonium fuel cycle requires. The thorium fuel cycle therefore combines safe reactors, a long-term source of abundant fuel, and no need for expensive fuel-enrichment facilities.
  • The molten-salt-fueled reactor operates much hotter than LWR reactors, from 650 °C in the tested MSRE (see above) and related designs, to as hot as 950 °C in untested designs. So, very efficient Brayton cycle (gas turbine) generators are possible. The MSRE already demonstrated operation at 650°C, making the MSR the most advanced of the "generation IV reactors." The efficiency from high temperatures reduces fuel use, waste emission and the cost of auxiliary equipment (major capital expenses) by 50% or more.
  • MSRs work in small sizes, as well as large, so a utility could easily build several small reactors (say 100 MWe) from income, reducing interest expense and business risks.
  • Molten salt fuel reactors are not experimental. Several have been constructed and operated at 650 °C temperatures for extended times, with simple, practical validated designs. There's no need for new science at all, and very little risk in engineering new, larger or modular designs.
  • In most new reactor designs, the longest-lead item is the safety testing of solid fuel elements. Fuel tests for a new reactor design usually must cover several three-year refueling cycles, and therefore take more than ten years. Since the MSR uses molten salts, the fuel has already been validated, and development can proceed more quickly.
  • The reactor, like all nuclear plants, has little effect on biomes. In particular, it uses only small amounts of land, relatively small amounts of construction, and the waste is separated from the biome, unlike both fossil and renewable energy projects.

Technological disadvantages

On-line reprocessing advantages

A molten salt reactor's fuel can be continuously reprocessed with a small adjacent chemical plant. Weinberg's groups at Oak Ridge National Laboratory found that a very small reprocessing facility can service a large 1 GW power plant: All the salt has to be reprocessed, but only every ten days. The reactor's total inventory of expensive, poisonous radioactive materials is therefore much smaller than in a conventional light-water-reactor's fuel cycle, which have to store spent fuel rod assemblies. Also, everything except fuel and waste stays inside the plant. The reprocessing cycle is:

Thorium cycle advantages

Actinides Half-life Fission products
244Cm 241Pu f 250Cf 243Cmf 10–30 y 137Cs 90Sr 85Kr
232 f 238Pu f is for
fissile
69–90 y 151Sm nc➔
4n 249Cf  f 242Amf 141–351 No fission product
has half-life 102
to 2×105 years
241Am 251Cf  f 431–898
240Pu 229Th 246Cm 243Am 5–7 ky
4n 245Cmf 250Cm 239Pu f 8–24 ky
233U    f 230Th 231Pa 32–160
4n+1 234U 4n+3 211–290 99Tc 126Sn 79Se
248Cm 242Pu 340–373 Long-lived fission products
237Np 4n+2 1–2 my 93Zr 135Cs nc➔
236U 4n+1 247Cmf 6–23 107Pd 129I
244Pu 80 my >7% >5% >1% >.1%
232Th 238U 235U    f 0.7–12by fission product yield

The thorium fuel cycle, like other breeder reactor fuel cycles with reprocessing, can potentially burn all the actinides that produce most of the radioactivity of spent nuclear fuel in the time range from several hundred years after fuel use, after the decay of the 30-year fission products cesium-137 and strontium-90, up to several hundred thousand years, when long-lived fission products like technetium-99 become significant. The open fuel cycle of the light-water reactors of the current nuclear power industry leaves substantial quantities of isotopes of plutonium and minor actinides in its spent fuel.

Reduction of radiation in this time range is dependent on nearly complete actinide removal and recycling during reprocessing. If even small amounts are not removed and instead are disposed as part of reprocessing wastes, much of the advantage is lost.

The thorium cycle produces lower levels of heavy actinides than the uranium-238 and plutonium cycle, because the mass of the starting point is lower, giving more opportunities for destruction by fission before reaching higher masses. However, the thorium cycle produces protactinium-231 (half-life 31,000 years) via (n,2n) reactions with fast neutrons. Both Pa-231 and the heavy actinides can be eventually destroyed in a closed fuel cycle via neutron capture and fission. However if a MSR chemically isolates Pa-233 outside the reactor core to avoid neutron capture, any Pa-231 will come with it, and continue to accumulate while the Pa-233 decays to U-233 and returns to the reactor.

Technological advantages

  • Control of the salt's corrosivity is easy. The uranium buffers the salt, forming more UF4 from UF3 as more fluorine is present. UF3 can be regenerated by adding small amounts of metallic beryllium to absorb F. In the MSRE, a beryllium rod was inserted into the salt until the UF3 was the correct concentration.[9]
  • Extensive validation (fuel rod design validation normally takes years and prevents effective deployment of new nuclear technologies) is not needed. The fuel is molten, chemical reprocessing eliminates reaction products, and there are tested fuel mixtures, notably FLi7BeU.
  • Molten-fuel reactors can be made to have passive nuclear safety: Tested fuel-salt mixtures have negative reactivity coefficients, so that they decrease power generation as they get too hot. Most fuel-salt reactor vessels also have a freeze plug at the bottom that has to be actively cooled. If the cooling fails, the fuel drains to a subcritical storage facility.
  • Continuous reprocessing simplifies numerous reactor design and operating issues. For example, the neutron absorbing effects of xenon-135 are not present. Neutron absorption from fission products is continuously mitigated. Transuranic elements, the long-lived "wastes" of light water reactors, are burned as fuel.
  • A fuel-salt reactor is mechanically and neutronically simpler than light-water reactors. There are only two items in the core: fuel salts and moderators. This reduces concerns with moderating interactions with positive void coefficients as water boils, chemical interactions, etc. (In fact since water is a moderator, boiling produces a stabilizing negative void coefficient in a thermal reactor)
  • Coolant and piping need never enter the high-neutron-flux zone, because the fuel is used to cool the core. The fuel is cooled in low-neutron-flux heat-exchangers outside the core. This reduces worries about neutron effects on pipes, testing, development issues, etc.
  • The salt distillation process means that chemical separation and recycling of fission products, say for nuclear batteries, are quite cheap. Xenon and other valuable transmuted noble gases separate out of the molten fuel in the pump-bowl. Any transuranic elements go right back into the fuel for burn-up.

Design challenges

Molten salt reactors, nevertheless, present a number of design challenges. Known issues include:

  • Graphite moderated, water-cooled, solid-fueled reactor designs can be susceptible to increases in reactivity with voids in their coolant (positive coolant void coefficient - if the reactor loses coolant, the reaction speeds up), making such designs unsafe. Unlike other reactors, however, in the single-fluid MSR, the fuel and the coolant are intimately mixed molten salts. So, if the MSR has an incident that causes voids in the coolant, this incident will also cause voids in the fuel, leading to cessation of the nuclear reaction. In addition, the coolant/fuel can be easily removed from the reactor within seconds by opening a valve below the reactor, and allowing gravity to push the molten salt into holding tanks set aside for this purpose, designed to store the salts in a non-critical configuration.
  • High neutron fluxes and temperatures in a compact MSR core can change the shape of a graphite moderator element, causing it to require refurbishing in as little as four years of operation. Eliminating graphite from sealed piping was a major incentive to switch to a single-fluid design.[10] Most MSR designs do not use graphite as a structural material, and arrange for it to be easy to replace. At least one design used graphite balls floating in salt, which could be removed and inspected continuously without shutting down the reactor.[11]
  • The high neutron density in the core rapidly transmutes lithium-6 to tritium, a radioactive isotope of hydrogen. In an MSR, the tritium forms hydrogen fluoride (HF). Tritium fluoride is a corrosive, chemically reactive, radioactive gas. Because of this, all MSR designs use lithium-7 isotope for their carrier salts in order to prevent tritium formation. The MSRE proved that lithium-6 removal from the fuel salt worked to prevent tritium formation. Since lithium-7 is at least 14% heavier than lithium-6, and represents the most common isotope of lithium, the lithium-6 is comparatively easy and inexpensive to extract from naturally occurring lithium in isotopic separation, as gaseous lithium diffuses at dramatically different speeds due to the difference in atomic weights. Vacuum distillation of lithium achieves efficiencies of up to 8% per stage and only requires heating of raw lithium in a vacuum chamber.
  • Some slow corrosion occurs even in the special nickel alloy, Hastelloy-N used for the reactor. The corrosion is faster if the reactor is exposed to hydrogen which forms corrosive HF gas. Exposure to water-vapor within the piping causes uptake of corrosive amounts of hydrogen, so practical MSRs operate the salt under a blanket of dry inert gas, usually helium.[12]
  • When cold, the fuel salts radiogenically produce corrosive, chemically reactive fluorine gas. Although a very slow process, the salts should be defueled and wastes removed before extended shutdowns to avoid (non-radioactive) fluorine gas production. Unfortunately, this was discovered the unpleasant way, while the MSRE was shut-down over a 20-year period.

An MSR based on chloride salts (e.g. sodium chloride as the carrier salt) has many of the same advantages. However, the heavier nuclei of chlorine are less moderating, which causes the reactor to be a fast reactor. Theoretically, it wastes even fewer neutrons and breeds more efficiently, though it may be less safe. It would require isotopically-pure chlorine-37, to avoid neutron activation of chlorine-35 into the long-lived radioactive activation product chlorine-36.

Fuel cycle concerns

  • There is no need for fuel fabrication. This greatly reduces the MSR's fuel expenses. It poses a business challenge, because reactor manufacturers customarily get their long-term profits from fuel fabrication. A government agency could, however, type-license a design, which utilities could replicate. Since it uses raw fuel, basically just a mixture of chemicals, current reactor vendors don't want to develop it. They derive their long-term profits from sales of fabricated fuel assemblies.
  • A safe thorium breeder reactor using slow thermal-energy neutrons also has a low breeding rate. Each year it can only breed thorium into about 109% of the U233 fuel it consumes. This means that obtaining enough U233 for a new reactor can take eight years or more, which would slow deployment of this type of reactor. Most practical, fast deployment plans would start the new thorium reactors with plutonium from existing light-water reactor wastes or decommissioned nuclear weapons. This scheme also decreases society's stock of high-level wastes.

Economical and social advantages

Combining the above, some form of molten-salt thorium breeder could be the most efficient well-developed energy source known, whether measured by cost per kW, capital cost or social costs.

  • Thorium's fuel cycle resists proliferation in two ways:
    • It is verifiable because the epithermal thorium breeder produces only at most 9% more fuel than it burns in each year. Building bombs quickly will take power plants out of operation.
    • Also, an easy variation of the thorium fuel cycle would contaminate the Th232 breeding material with chemically inseparable Th230. The Th230 breeds into U232, which has a powerful gamma-ray emitter in its decay chain (Thallium-208) that makes the reactor fuel U233/U232 impractical in a bomb, because it harms electronics.
  • The Earth's crust contains about three times as much Thorium as uranium-238, or 400 times as much as uranium-235, which makes it about as abundant as lead.
  • Thorium is cheap. Currently, it costs US$ 30/kg. In the 2000s, the price of uranium has risen above $100/kg, not including the cost of enrichment, and fuel element fabrication.

Molten-salt cooled reactors

Molten-salt-fueled reactors are quite different from molten-salt-cooled solid-fuel reactors, called simply "Molten Salt Reactor System" in the Generation IV proposal, also called MSCR, which is also the acronym for the Molten Salt Converter Reactor design. It cannot reprocess fuel easily and has fuel rods that need to be fabricated and validated, delaying deployment by up to twenty years from project inception. However, since it uses fabricated fuel, reactor manufacturers can still profit by selling fuel assemblies.

The MSCR retains the safety and cost advantages of a low-pressure, high-temperature coolant, also shared by liquid metal cooled reactors. Notably, there's no steam in the core to cause an explosion, and no large, expensive steel pressure vessel. Since it can operate at high temperatures, the conversion of the heat to electricity can also use an efficient, lightweight Brayton cycle gas turbine.

Much of the current research on MSCRs is focused on small compact heat exchangers. By using smaller heat exchangers, less molten salt needs to be used and therefore significant cost savings could be achieved.

Molten salts can be highly corrosive, more so as temperatures rise. For the primary cooling loop of the MSR, a material is needed that can withstand corrosion at high temperatures and intense radiation. Experiments show that Hastelloy-N and similar alloys are quite suited to the tasks at operating temperatures up to about 700 °C. However, long-term experience with a production scale reactor has yet to be gained. Higher operating temperatures would be desirable, especially since at 850 °C thermo chemical production of hydrogen becomes possible. Materials for this temperature range have not yet been found, though carbon composites, carbides, and refractory metal based or ODS alloys might be feasible.

Fused salt selection

Molten FLiBe

The salt mixtures are chosen to make the reactor safer and more practical. Fluorides are favored because fluorine doesn't need expensive isotope separation (as chlorine does). It does not easily become radioactive under neutron bombardment. It also absorbs fewer neutrons and slows ("moderates") neutrons better. Low-valence fluorides boil at high temperatures, though many pentafluorides and hexafluorides boil at low temperatures. They also must be very hot before they break down into simpler compounds, or corrode materials (they are "chemically stable").

Reactor salts are also eutectic mixtures to reduce their melting point. This makes a heat engine more efficient, because more heat can be removed from the salt before reheating it in the reactor.

Some salts are so useful that isotope separation is worthwhile. Chlorides permit fast breeder reactors to be constructed using molten salts. Not nearly as much work has been done on reactor designs using them. Chlorine must be purified to Cl37 to reduce production of radioactive elements. Also, any lithium in a salt mixture must be purified lithium-7 to reduce tritium production.

Due to the high "redox window" of fused fluoride salts, the chemical potential of the fused salt system can be changed. Fluorine-Lithium-Beryllium ("FLiBe") can be used with beryllium additions to lower the electrochemical potential and almost eliminate corrosion. However, beryllium is extremely toxic. Many other salts can cause corrosion, especially if the reactor is hot enough to make hydrogen.

To date, most research has focused on FLiBe, because Lithium and Beryllium are reasonably effective moderators, and form a eutectic salt mixture with a lower melting point than each of the constituent salts. Beryllium also performs neutron doubling, improving the neutron economy. This process occurs when the Beryllium nucleus re-emit two neutrons after absorbing a single neutron. For the fuel carrying salts, generally 1% or 2% (by mole) of UF4 is added. thorium and plutonium fluorides have also been used. The MSFR is the only system that has run a single reactor, the MSRE, from all three known nuclear fuels.

Material Total neutron capture
relative to graphite
(per unit volume)
Moderating ratio
(Avg. 0.1 to 10 eV)
Heavy water 0.2 11449
Light water 75 246
Graphite 1 863
Sodium 47 2
UCO 285 2
UO2 3583 0.1
2LiF–BeF2 8 60
LiF–BeF2–ZrF4 (64.5–30.5–5) 8 54
NaF–BeF2 (57–43) 28 15
LiF–NaF–BeF2 (31–31–38) 20 22
LiF–ZrF4 (51–49) 9 29
NaF–ZrF4 (59.5–40.5) 24 10
LiF-NaF–ZrF4 (26–37–37) 20 13
KF–ZrF4 (58–42) 67 3
RbF–ZrF4 (58–42) 14 13
LiF–KF (50–50) 97 2
LiF–RbF (44–56) 19 9
LiF–NaF–KF (46.5–11.5–42) 90 2
LiF–NaF–RbF (42–6–52) 20 8

Above is a table comparing the neutron capture and moderating efficiency of several materials. Red are Be bearing salts, blue are ZrF4 bearing salts, and green are LiF bearing salts. (Source: ORNL/TM-2005/218, Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR), December 2005, D. T. Ingersoll)

Fused salt purification and reprocessing

Salts must be extremely pure initially, and would most likely be continuously cleaned in a large-scale molten salt reactor. Any water vapor in the salt will form hydrofluoric acid (HF) which is extremely corrosive. Other impurities can cause non-beneficial chemical reactions and would most likely have to be cleansed from the system. It should be noted that most power plants have to ensure that the primary coolant they are using is extremely pure; otherwise, they would encounter corrosion issues as well.

The possibility of online reprocessing can be an advantage of the MSR design. Continuous reprocessing ensures a low inventory of fission products at all times, which improves neutron economy. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. To allow breeding from thorium, the intermediate product protactinium-233 has to be removed from the reactor and stored for some months while it decays into uranium 233. Left in the fuel it would absorb too many neutrons to make breeding with a graphite moderator and thermal spectrum possible (though with alternate designs in which the thorium is kept in a separate fluid from the fuel, the protactinium can simply be diluted with a larger volume of thorium fluid which proportionately reduces the neutron absorption; also some heavy water moderated reactor designs could overcome this, albeit at a lower thermal efficiency). The necessary reprocessing technology, which has to process the complete fuel every 10 days, has only been demonstrated at laboratory scale. For a power reactor such a large reprocessing facility is currently deemed uneconomic.

Political issues

To exploit the molten salt reactor's breeding potential to the fullest, the reactor must be co-located with a reprocessing facility. Nuclear reprocessing does not occur in the U.S. because no commercial provider is willing to undertake it. The regulatory risk and associated costs are very great because the regulatory regime has varied dramatically in different administrations. [13] UK, France, Japan, Russia and India currently operate some form of fuel reprocessing.

Some U.S. Administration departments have feared that fuel reprocessing in any form could pave the way to the plutonium economy with its associated proliferation dangers.[14]

A similar argument led to the shutdown of the Integral Fast Reactor project in 1994. [15] The proliferation risk for a thorium fuel cycle stems from the potential separation of U-233, which might be used in nuclear weapons, though only with considerable difficulty.

Comparison to ordinary light water reactors

MSRs can be safer. Molten salts trap fission products chemically, and react slowly or not at all in air. Also, the fuel salt does not burn in air or water. The core and primary cooling loop is operated at near atmospheric pressure, and has no steam, so a pressure explosion is impossible. Even in the unlikely case of an accident, most radioactive fission products would stay in the salt instead of dispersing into the atmosphere. A molten core is meltdown-proof, so the worst possible accident would be a leak. In this case, the fuel salt can be drained into passively cooled storage, managing the accident. Neutron-producing accelerators have even been proposed for some super-safe subcritical experimental designs.

Some types of molten salt reactors are very efficient. Since the core and primary coolant loop are low pressure, it can be constructed of thin, relatively inexpensive weldments. So, it can be far less expensive than the massive pressure vessel required by the core of a light water reactor. Also, some form of fluid-fueled thorium breeder could use less fissile material per megawatt than any other reactor. Molten salt reactors can run at extremely high temperatures, with extremely high efficiencies when producing electricity. The temperatures are high enough to produce process heat for hydrogen production or other chemical reactions. Because of this, they have been included in the GEN-IV roadmap for further study.

Molten-salt-fueled thorium breeders close the nuclear fuel cycle and potentially eliminate the need for both fuel enrichment and fuel fabrication, both major expenses. The Liquid Fluoride Thorium Reactor or LFTR is an example of this technology approach.

The MSR also has far better neutron economy and, depending on the design, a harder neutron spectrum than conventional light water reactors. So, it can operate with less reactive fuels. Some designs (such as the MSRE) can operate a single design from all three common nuclear fuels. For example, it can breed from uranium-238, thorium or even burn the transuranic spent nuclear fuel from light water reactors. In contrast, a water-cooled reactor cannot completely consume the plutonium it produces, because the increasing impurities from the fission wastes capture too many neutrons, "poisoning" the reaction.

Molten salt-fueled thorium breeders can operate for extended periods, possibly decades, without refueling, by chemically precipitating neutronic poisons.

MSRs scale over a wide range of powers. Reactors as small as several megawatts have been constructed and operated. Theoretical designs up to several gigawatts have been proposed[16].

Because of their lightweight structures and compact cores, MSRs weigh less per watt (that is, they have a greater "specific power") than other proven reactor designs. So, in small sizes, with long refueling intervals, they are an excellent choice to power vehicles, including ships, aircraft and spacecraft.

Commercialization

On Aug. 16, 2006 the North-American Energy Group Corporation has announced its intention to research, and develop Thorium-based nuclear power generation facilities, and Thorium-based power cells.[17]

See also

References

  1. ^ Safety and Licensing Aspects of the Molten Salt Reactor, page 15 (last), Charles Forsberg, Oak Ridge National Laboratory
  2. ^ Section 5.3, WASH 1097, Energy From Thorium's Document Repository "The Use of Thorium in Nuclear Power Reactors", For sale by the Superintendent of Documents, U.S., Washington, DC, or available in PDF from , Accessed 11/23/09
  3. ^ Section 2.5.2, WASH 1097, Energy From Thorium's Document Repository "The Use of Thorium in Nuclear Power Reactors", For sale by the Superintendent of Documents, U.S., Washington, DC, or available in PDF from , Accessed 11/23/09
  4. ^ Fuji MSR pp. 821-856, Jan 2007 20MB PDF
  5. ^ FUJI Reactor, in the MSR article of the Encyclopedia of the Earth
  6. ^ Fuji Molten salt reactor, December 19, 2007
  7. ^ Fuji Molten Salt reactor, Ralph Moir Interviews and other nuclear news, March 19, 2008
  8. ^ Energy from Thorium blog->Reactor Design->Graphite and Two-Fluid vs. One-Fluid LFRs Viewed 6/2007
  9. ^ Briant, Raymond C.; Alvin M. Weinberg (1957). "Molten Fluorides as Power Reactor Fuels" (PDF). Nuclear Science and Engineering. pp. 2, 797–803. http://www.energyfromthorium.com/pdf/NSE_moltenFluorides.pdf. Retrieved 2008-05-18.  
  10. ^ Energy from Thorium blog->Reactor Design->Graphite and Two-Fluid vs. One-Fluid LFRs Viewed 6/2007
  11. ^ ORNL-4548: Molten-Salt Reactor Program: Semiannual Progress Report for Period Ending February 28, 1970, pg. 57
  12. ^ W.D. Manely et al. (1960). Metallurgical Problems in Molten Fluoride Systems. Progress in Nuclear Energy, Vol. 2, pg. 164-179
  13. ^ Congressional Research Service Report RS22542, March 27, 2008, Accessed 11/23/09
  14. ^ Congressional Research Service Report RS22542, See Presidents H. Bush and Carter
  15. ^ Sustainable Nuclear's Reprint of Charles Till's article series about the IFR, published in The Republic, June-September, 2005 See part 3, "The Shutdown of the IFR," which fingers John Kerry as the main opponent, and whose main concern was said to be the proliferation dangers of a plutonium economy. This is mistaken in the case of the IFR, whose plutonium would never be purified, and would normally never leave the power plant.
  16. ^ Weinberg et al. WASH 1080, ORNL
  17. ^ "Northamerican Energy Group to Launch Thorium Power Generating Research Development Division". GlobeNewswire. 2006-08-16. http://www.globenewswire.com/newsroom/news.html?d=103968. Retrieved 2009-05-16.  

Further reading

External links


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