Molten salt reactor
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.
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 (1954) was primarily motivated by the small size that the design could provide, while the Molten-Salt Reactor Experiment (1965–1969) was a prototype for a thorium fuel cycle breeder reactor nuclear power plant. The increased research into Generation IV reactor designs included a renewed interest in the technology.
- 1 History
- 2 Twenty-first century
- 3 Variants
- 4 Molten-salt fueling options
- 5 Molten-salt-cooled reactors
- 6 Fused salt selection
- 7 Fissile fuel reprocessing issues
- 8 Comparison to light water reactors
- 9 See also
- 10 References
- 11 Further reading
- 12 External links
Aircraft reactor experiment
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.
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.
Molten-salt reactor experiment
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. Despite the success, the MSR program closed down in the early 1970s in favor of the liquid metal fast-breeder reactor (LMFBR), after which research stagnated in the United States. As of 2011[update], the ARE and the MSRE remained the only molten-salt reactors ever operated.
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.
- 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.
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. 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. 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.
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. They also researched the option of helium gas as an alternative coolant.
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.
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. A reduction in activity occurred after 1986 due to the Chernobyl accident, along with a general stagnation of nuclear power and the nuclear industry.(p381)
Terrestrial Energy Inc. (TEI), a Canadian based company, is developing a DMSR design called the Integral Molten Salt Reactor (IMSR). The IMSR is designed to be deployable as a small modular reactor (SMR) and will be constructed in three configurations ranging from 80 to 600 MW. With high operating temperatures, the IMSR has applications in industrial heat markets as well as traditional power markets. The main design features include neutron moderation from graphite, fueling with low-enriched uranium and a compact and replaceable Core-unit. The latter feature permits the operational simplicity necessary for industrial deployment.
Under Jiang Mianheng's direction, China initiated a thorium molten-salt reactor research project. It was formally announced at the Chinese Academy of Sciences (CAS) annual conference in January 2011. A 100-MW demonstrator of the solid fuel version (TMSR-SF), based on pebble bed technology, was to be ready by 2024. A 10-MW pilot and a larger demonstrator of the liquid fuel (TMSR-LF) variant are targeted for 2024 and 2035 respectively.
Seaborg Technologies, a company based in Denmark, is developing the core for a Molten Salt Waste-burner (MSW). The MSW is a high temperature, single salt, thermal MSR designed to go critical on a combination of thorium and nuclear waste from conventional nuclear reactors. The MSW design is modular. The reactor core is estimated to be replaced every 6–10 years. However, the fuel will not be replaced and will burn for the entire power plant lifetime. The first version of the Seaborg core is planned to produce 50 MWth power and could consume approximately 1 ton (not considering natural decays) of transuranic waste over its 60 years power plant lifetime. After 60 years the 233U concentration in the fuel salt is high enough to initiate a closed thorium fuel cycle in the next generation power plant.
The CNRS project EVOL (Evaluation and viability of liquid fuel fast reactor system) project, with the objective of proposing a design of the MSFR (Molten Salt Fast Reactor), released its final report in 2014. The various molten salt reactor projects like FHR, MOSART, MSFR, and TMSR have common themes in basic R&D areas, according to a 2014 paper giving an overview of the MSR in a GenV context. Another paper gives an overview of the MSFR. More resources are available in the MSFR bibliography.
Ratan Kumar Sinha, Chairman of Atomic Energy Commission of India, stated in 2013: "India is also investigating Molten Salt Reactor (MSR) technology. We have molten salt loops operational at BARC."
The FUJI MSR is a 100 to 200 MWe LFTR, using technology similar to the Oak Ridge project. A consortium including members from Japan, the U.S. and Russia are developing the project. The project would likely take 20 years to develop a full size reactor, but the project seems to lack funding.
The Alvin Weinberg Foundation is a British non-profit organization founded in 2011, dedicated to raising awareness about the potential of thorium energy and LFTR. It was formally launched at the House of Lords on 8 September 2011. It is named after American nuclear physicist Alvin M. Weinberg, who pioneered thorium molten salt reactor research.
A study on MSRs completed in July 2015 by Energy Process Developments, funded by Innovate UK, summarizes MSR activity internationally. It looks at the feasibility of developing a pilot scale demonstration MSR in the UK. A review of potential UK sites is given along with an insight into the UK regulatory process for innovative reactor technology. The technical review of six MSR designs led to the selection of the Stable Salt Reactor, designed by Moltex Energy, as most suitable for UK implementation.
Kirk Sorensen, former NASA scientist and chief nuclear technologist at Teledyne Brown Engineering, is a long-time promoter of the thorium fuel cycle, coining the term liquid fluoride thorium reactor. In 2011, Sorensen founded Flibe Energy, a company aimed at developing 20–50 MW LFTR reactor designs to power military bases. (It is easier to approve novel military designs than civilian power station designs in today's US nuclear regulatory environment).
Transatomic Power was created by Ph.D. students from MIT including CEO Leslie Dewan and Mark Massie, and Russ Wilcox of E Ink. They are pursuing what they term a Waste-Annihilating Molten Salt Reactor (acronym WAMSR), intending to consume existing spent nuclear fuel. Transatomic received venture capital funding in early 2015.
In January 2016, the United States Department of Energy announced a $80m award fund to develop Generation IV reactor designs. One of the two beneficiaries, Southern Company will use the funding to develop a Molten Chloride Fast Reactor (MCFR), a type of MSR developed earlier by British scientists.
Liquid-salt very-high-temperature reactor
As of September 2010[update], research was continuing for reactors that utilize molten salts for coolant. Both the traditional molten-salt reactor and the very high temperature reactor (VHTR) were selected as potential designs for study under the Generation Four Initiative (GEN-IV). One version of the VHTR under study was 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 salt candidates is >1400 °C); low-pressure cooling that can be used to more easily match hydrogen production facility conditions (most thermochemical 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. This concept is now referred to as "fluoride salt-cooled high-temperature reactor" (FHR).
Liquid Fluoride Thorium Reactor (LFTR)
Reactors containing molten thorium salt, called liquid fluoride thorium reactors (LFTR), would tap the abundant energy source of the thorium fuel cycle. Private companies from Japan, Russia, Australia and the United States, and the Chinese government, have expressed interest in developing this technology.
Advocates estimate that five hundred metric tons of thorium could supply all U.S. energy needs for one year. The U.S. Geological Survey estimates that the largest known U.S. thorium deposit, the Lemhi Pass district on the Montana-Idaho border, contains thorium reserves of 64,000 metric tons.
Molten-salt fueling options
The LFTR design was strongly supported by Alvin Weinberg, who patented the light-water reactor and was a director of the U.S.'s Oak Ridge National Laboratory. In 2016 Nobel prize winning physicist Carlo Rubbia, former Director General of CERN, claimed that one of the main reasons why research was cut is that thorium is difficult to turn into a nuclear weapon.
Molten-salt-cooled solid-fuel reactors are quite different from molten-salt-fueled reactors. They are called "molten salt reactor system" in the Generation IV proposal, also called Molten Salt Converter Reactor (MSCR). These reactors were additionally referred to as advanced high-temperature reactors (AHTRs), but since about 2010 the preferred DOE designation is fluoride high-temperature reactors (FHR).
The FHR concept 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 FHR retains the safety and cost advantages of a low-pressure, high-temperature coolant, also shared by liquid metal cooled reactors. Notably, steam is not created in the core (as is present in BWRs), and no large, expensive steel pressure vessel (as required for PWRs). Since it can operate at high temperatures, the conversion of the heat to electricity can use an efficient, lightweight Brayton cycle gas turbine.
Molten salts can be highly corrosive and corrosivity increases with temperature. For the primary cooling loop, a material is needed that can withstand corrosion at high temperatures and intense radiation. Experiments show that Hastelloy-N and similar alloys are suited to these tasks at operating temperatures up to about 700 °C. However, operating experience is limited. Still higher operating temperatures are desirable – at 850 °C thermochemical production of hydrogen becomes possible. Materials for this temperature range have not been validated, though carbon composites, molybdenum alloys (e.g. TZM), carbides, and refractory metal based or ODS alloys might be feasible.
Fused salt selection
The salt mixtures are chosen to make the reactor safer and more practical. Fluoride salts are favored, because fluorine has only one stable isotope (F-19), and does not easily become radioactive under neutron bombardment. Both of these make fluorine better than chlorine, which has two stable isotopes (Cl-35 and Cl-37), as well as a slow-decaying isotope between them which facilitates neutron absorption by Cl-35. Compared to chlorine and other halides, fluorine 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 their constituent elements. Such molten salts are "chemically stable" when maintained well below their boiling points.
On the other hand, some salts are so useful that isotope separation of the halide is worthwhile. Chlorides permit fast breeder reactors to be constructed using molten salts. Much less research has been done on reactor designs using chloride salts. Chlorine, unlike fluorine, must be purified to isolate the heavier stable isotope, chlorine-37, thus reducing production of sulfur tetrafluoride that occurs when chlorine-35 absorbs a neutron to become chlorine-36, then degrades by beta decay to sulfur-36.
Similarly, any lithium present in a salt mixture must be in the form of purified lithium-7, because lithium-6 effectively captures neutrons and produces tritium. Even if pure 7Li is used, salts containing lithium will cause significant tritium production, comparable with heavy water reactors.
Reactor salts are usually close to eutectic mixtures to reduce their melting point. A low melting point simplifies melting the salt at startup and reduces the risk of the salt freezing as it is cooled in the heat exchanger.
Due to the high "redox window" of fused fluoride salts, the redox potential of the fused salt system can be changed. Fluorine-Lithium-Beryllium ("FLiBe") can be used with beryllium additions to lower the redox potential and almost eliminate corrosion. However, since beryllium is extremely toxic, special precautions must be engineered into the design to prevent its release into the environment. Many other salts can cause plumbing corrosion, especially if the reactor is hot enough to make highly reactive 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-emits 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.
|Material||Total neutron capture
relative to graphite
(per unit volume)
(Avg. 0.1 to 10 eV)
|ZrH||~0.2||~0 if <0.14 eV, ~11449 if >0.14 eV|
Fused salt purification
Techniques for preparing and handling molten salt were first developed at Oak Ridge National Lab. The purpose of salt purification was to eliminate oxides, sulfur and metal impurities. Oxides could result in the deposition of solid particles in reactor operation. Sulfur had to be removed because of its corrosive attack on nickel-based alloys at operational temperature. Structural metal such as chromium, nickel, and iron had to be removed for corrosion control.
A water content reduction purification stage using HF and helium sweep gas was specified to run at 400 °C. Oxide and sulfur contamination in the salt mixtures were removed using gas sparging of HF – H2 mixture, with the salt heated to 600 °C.(p8) Structural metal contamination in the salt mixtures were removed using hydrogen gas sparging, at 700 °C.(p26) Solid ammonium hydrofluoride was proposed as a safer alternative for oxide removal.
Fused salt processing
The possibility of online processing can be an MSR advantage. Continuous processing would reduce the inventory of fission products, control corrosion and improve neutron economy by removing fission products with high neutron absorption cross-section, especially xenon. This makes the MSR particularly suited to the neutron-poor thorium fuel cycle. Online fuel processing can introduce risks of fuel processing accidents,(p15) which can trigger release of radio isotopes.
In some thorium breeding scenarios, the intermediate product protactinium-233 would be removed from the reactor and allowed to decay into highly pure uranium-233, an attractive bomb-making material. More modern designs propose to use a lower specific power or a separate large thorium breeding blanket. This dilutes the protactinium to such an extent that few protactinium atoms absorb a second neutron or, via a (n, 2n) reaction (in which an incident neutron is not absorbed but instead knocks a neutron out of the nucleus), generate uranium-232. Because U-232 has a short half-life and its decay chain contains hard gamma emitters, it makes the isotopic mix of uranium less attractive for bomb-making. This benefit would come with the added expense of a larger fissile inventory or a 2-fluid design with a large quantity of blanket salt.
The necessary fuel salt reprocessing technology has been demonstrated, but only at laboratory scale. A prerequisite to full-scale commercial reactor design is the R&D to engineer an economically competitive fuel salt cleaning system.
Fissile fuel reprocessing issues
Reprocessing refers to the chemical separation of fissionable uranium and plutonium from spent nuclear fuel. The recovery of uranium or plutonium could increase the risk of nuclear proliferation. In the United States the regulatory regime has varied dramatically in different administrations.
In the original 1971 Molten Salt Breeder Reactor proposal, uranium reprocessing was scheduled every ten days as part of reactor operation.(p181) Subsequently a once-through fueling design was proposed that limited uranium reprocessing to every thirty years at the end of useful salt life.(p98) A mixture of uranium-238 was called for to make sure recovered uranium would not be weapons grade. This design is referred to as denatured molten salt reactor. If reprocessing were to be prohibited then the uranium would be disposed with other fission products.
Comparison to light water reactors
MSRs, especially those with the fuel dissolved in the salt differ considerably from conventional reactors. Reactor core pressure can be low and the temperature much higher. In this respect an MSR is more similar to a liquid metal cooled reactor than to a conventional light water cooled reactor. MSRs are often planned as breeding reactors with a closed fuel cycle – as opposed to the once-through fuel currently used in U.S. nuclear reactors.
Safety concepts rely on a negative temperature coefficient of reactivity and a large possible temperature rise to limit reactivity excursions. As an additional method for shutdown, a separate, passively cooled container below the reactor can be included. In case of problems and for regular maintenance the fuel is drained from the reactor. This stops the nuclear reaction and acts as another second cooling system. Neutron-producing accelerators have been proposed for some super-safe subcritical experimental designs.
Cost estimates from the 1970s were slightly lower than for conventional light-water reactors.
The temperatures of some proposed designs are high enough to produce process heat for hydrogen production or other chemical reactions. Because of this, they are included in the GEN-IV roadmap for further study.
MSR offers many potential advantages over current light water reactors:
- Inherently safe design (safety by passive components and the strong negative temperature coefficient of reactivity of some designs). In some designs, the fuel and the coolant are the same fluid, so a loss of coolant removes the reactor's fuel. Unlike steam, fluoride salts dissolve poorly in water, and do not form burnable hydrogen. Unlike steel and solid uranium oxide, molten salts are not damaged by the core's neutron bombardment.
- A low-pressure MSR lacks a LWR's high-pressure radioactive steam and therefore do no experience leaks of radioactive steam and cooling water, and the expensive containment, steel core vessel, piping and safety equipment needed to contain radioactive steam.
- MSRs make closed nuclear fuel cycles cheaper and more practical. If fully implemented, a closed nuclear fuel cycle reduces environmental impacts: The chemical separation makes long-lived actinides back into reactor fuel. The discharged wastes are mostly fission products (nuclear ashes) with short half-lives. This reduces the needed geologic containment to 300 years rather than the tens of thousands of years needed by a light-water reactor's spent nuclear fuel. It also permits society to use more-abundant nuclear fuels.
- The fuel's liquid phase might be pyroprocessed to separate fission products (nuclear ashes) from actinide fuels. This may have advantages over conventional reprocessing, though much development is still needed.
- Fuel rods are not required.
- In new solid-fueled reactor designs, the longest-lead item is the safety testing of fuel element designs. Fuel tests usually must cover several three-year refueling cycles. However, several molten salt fuels have already been validated.
- Some designs can "burn" problematic transuranic elements from traditional solid-fuel nuclear reactors.
- An MSR can react to load changes in less than 60 seconds (unlike "traditional" solid-fuel nuclear power plants that suffer from xenon poisoning).
- Molten salt reactors can run at high temperatures, yielding high production efficiency. This reduces the size, expense and environmental impacts of a power plant.
- MSRs can offer a high "specific power," that is high power at a low mass as demonstrated by the ARE. Simplified MSR power plants may be suitable for ships.
- A possibly good neutron economy makes the MSR attractive for the neutron poor thorium fuel cycle.
- LWR's (and most other solid-fuel reactors) have no fundamental "off switch", but once the initial criticality is overcome, an MSR is comparatively easy and fast turn to off by letting the freeze plug melt.
- Little development compared to most Gen IV designs .
- Required onsite chemical plant to manage core mixture and remove fission products.
- Required regulatory changes to deal with radically different design features.
- MSR designs rely on nickel-based alloys to hold the molten salt. Alloys based on nickel and iron are prone to embrittlement under high neutron flux.(p83)
- Corrosion risk.
- As a breeder reactor, a modified MSR might be able to produce weapons-grade nuclear material.
- The MSRE and aircraft nuclear reactors used enrichment levels so high that they approach the levels of nuclear weapons. These levels would be illegal in most modern regulatory regimes for power plants. Some modern designs avoid this issue.
- Neutron damage to solid moderator materials can limit the core lifetime of an MSR that makes moderately fast neutrons. For example, the MSRE was designed so that its graphite moderator sticks had very low tolerances, so neutron damage could change their size without damage. "Two fluid" MSR designs are unable to use graphite piping because graphite changes size when it is bombarded with neutrons, and graphite pipes would crack and leak.
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- Ingersoll, D. T. (December 2005). "ORNL/TM-2005/218, Status of Physics and Safety Analyses for the Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR)" (PDF). ORNL. Retrieved 13 May 2010.
- Baron, Matthias; Böck, Helmuth; Villa, Mario. "TRIGA Reactor Characteristics". IAEA Education and Training. IAEA. Retrieved 2 June 2016.
- Gylfe, J.D. "US Patent 3,145,150, Aug. 18, 1954, Fuel Moderator Element for a Nuclear Reactor, and Method of Making". U.S. Patent Office. U.S. Government. Retrieved 2 June 2016.
- Massie, Mark; Dewan, Leslie C. "US 20130083878 A1, April 4, 2013, NUCLEAR REACTORS AND RELATED METHODS AND APPARATUS". U.S. Patent Office. U.S. Government. Retrieved 2 June 2016.
- Shaffer, J. H. (January 1971), Preparation and Handling of Salt Mixtures for the Molten Salt Reactor Experiment (PDF), ORNL-4616, Oak Ridge National Laboratory
- Ignatiev, Victor (1 April 2010). Critical issues of nuclear energy systems employing molten salt fluorides (PDF). Lisbon, Portugal: ACSEPT.
- CForsberg, Charles (June 2004). "Safety and Licensing Aspects of the Molten Salt Reactor" (PDF). 2004 American Nuclear Society Annual Meeting. Pittsburgh, Pennsylvania: American Nuclear Society.
- Andrews, Anthony (27 March 2008), "Nuclear Fuel Processing: U.S. Policy Development" (PDF), CRS Report for Congress, Congressional Research Service, RS22542
- Rosenthal, M.; Briggs, R.; Haubenreich, P., Molten-Salt Reactor Program: Semiannual Progress Report for Period Ending August 31, 1971 (PDF), ORNL-4728, Oak Ridge National Laboratory
- J. R. Engel; et al. (1980). "Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling" (PDF). ORNL/TM-7207. Oak Ridge National Lab, TN.
- LeBlanc, D. (2010) Denatured Molten Salt Reactors (DMSR): An Idea Whose Time Has Finally Come?. 31st Annual conference of the Canadian Nuclear Society & 34th CNS/CNA student conference, Vol. 2 of 2 ( 24–27 May 2010) Montreal, Quebec, Canada. ISBN 978-1-61782-363-3
- Plutonium(TRU) Transmutation and 233U Production by Single-Fluid Type Accelerator Molten-Salt Breeder (AMSB) Kazuo Furukawa, Yoshio Kato, Sergey E. Chigrinov, Int. Conf. Accelerator-driven Transmutation, Tech. Appl. (Las Vegas, 25–29 July 1994)
- Moir, M. W. (2002). "Cost of Electricity from Molten Salt Reactors (MSR)" (PDF). 138. Nuclear Technology: 93–95.
- US DOE Nuclear Energy Research Advisory Committee (2002). "A Technology Roadmap for Generation IV Nuclear Energy Systems" (PDF). GIF-002-00.
- Finnish research network for generation four nuclear energy systems. vtt.fi
- "Is the "Superfuel" Thorium Riskier Than We Thought?". Popular Mechanics. 5 December 2012.
- "Transatomic Power White Paper, v1.0.1, section 1.2" (PDF). Transatomic Power Inc. Retrieved 2 June 2016.
- Energy from Thorium's Document Repository Contains scanned versions of many of the U.S. government engineering reports, over ten thousand pages of construction and operation experience. This repository is the main reference for the aircraft reactor experiment and molten-salt fueled reactor's technical discussion.
- Weinberg, Alvin M. (1994). The First Nuclear Era: The Life and Times of a Technological Fixer. Springer Science & Business Media. ISBN 978-1-56396-358-2.
- Bruce Hoglund's Eclectic Interests Home Page Nuclear Power, Thorium, Molten Salt reactors, etc.
- Generation IV International Forum MSR website
- INL MSR workshop summary
- "Molten Salt Chemistry Plays a Prominant (sic) Role in Accelerator-Driven Transmutation Systems". Archived from the original on 23 May 2014.
- Material Considerations for Molten Salt Accelerator-based Plutonium Conversion Systems J.H. Devan et al.
|Wikimedia Commons has media related to Molten salt reactors.|
- International Thorium Energy Organisation – www.IThEO.org
- Idaho National Laboratory Molten Salt Reactor Fact Sheet
- Energy from Thorium Blog / Website
- Google TechTalks – Liquid Fluoride Thorium Reactor: What Fusion Wanted To Be by Dr. Joe Bonometti NASA / Naval Postgraduate School YouTube
- Pebble Bed Advanced High Temperature Reactor
- Thorium Remix LFTR in 5 Minutes and other LFTR Documentaries.
- Kun Chen from Chinese Academy of Sciences on China Thorium Molten Salt Reactor TMSR Program
- Review of Molten Salt Reactor Technology