The Molten Salt Reactor (MSR) system produces fission power from a molten salt fuel circulating in a fast or epithermal-spectrum reactor and contains an integrated fuel cycle.

In a Molten Salt Reactor (MSR), the fuel is dissolved in a fluoride salt coolant. Prior MSRs were mainly considered as thermal-neutron-spectrum graphite-moderated concepts. Since 2005 R&D has focused on the development of fast-spectrum MSR concepts (MSFR) combining the generic assets of fast neutron reactors (extended resource utilization, waste minimization) to those relating to molten salt fluorides as fluid fuel and coolant (favourable thermal-hydraulic properties, high boiling temperature, optical transparency). In addition, MSFRs exhibit large negative temperature and void reactivity coefficients, a unique safety characteristic not found in solid-fuel fast reactors (Mathieu et al, 2009). MSFR systems have been recognized as a long term alternative to solid-fuelled fast neutron systems with unique potential (negative feedback coefficients, smaller fissile inventory, easy in-service inspection, simplified fuel cycle, etc.).

Apart from MSR systems, other advanced reactor concepts are being studied employing liquid salt technology as primary coolant in Fluoride-cooled High-temperature Reactor (FHR), or intermediate coolant as an alternative to secondary sodium in Sodium Fast Reactors (SFR) and to intermediate helium in Very High Temperature Reactors (VHTR).

More generally speaking, the development of higher temperature salts as coolants could bring new nuclear and non-nuclear applications. These salts could facilitate heat transfer for nuclear hydrogen production concepts, concentrated solar electricity generation, oil refineries and shale oil processing facilities amongst other applications (Forsberg et al., 2007).

Fluoride-cooled High-temperature Reactors (FHRs) combine the use of liquid fluoride salt coolants (like MSRs), pool type cores and vessel configurations in common with many sodium reactor designs, and coated particle fuels similar to high temperature gas-cooled reactors (Forsberg et al., 2008). The two most developed FHR designs are the 1200 MWe Advanced High Temperature Reactor (AHTR) that employs prismatic fuel elements and the 410MWe Pebble Bed Advanced High Temperature Reactor (PB-AHTR). The better fluoride salt heat transport characteristics, as compared to helium, enable power densities 4 to 8 times greater as well as power levels over 4000 MWt with passive safety systems. Fuel cycle characteristics are essentially identical to those of the VHTR, while intermediate heat transport, power conversion and balance of plant are essentially identical to those of the “reference” MSR.

This diagram illustrates a schematic concept of the reactor system and does not represent the reference design.

Diagram source: http://www.ne.doe.gov/genIV/documents/gen_iv_roadmap.pdf

Advantages and challenges

The renewal and diversification of interests in molten salts led the MSR PSSC in 2008 to shift the R&D aims and objectives promoted in the original Generation IV Roadmap, issued in 2002, in order to include in a consistent body the different applications then envisioned for fuel and coolant salts.

Since then, two baseline concepts are considered which have large commonalities in basic R&D areas, particularly for liquid salt technology and materials behaviour (mechanical integrity, corrosion). These are:

• The MSFR system operated in the thorium fuel cycle. Although its potential has been assessed, specific technological challenges remain and the safety approach has to be established.
• The FHR system, a high temperature reactor with better compactness than the VHTR and passive safety potential for medium to very high unit power (> 2400 MWt).

In addition, opportunities offered by liquid salts for intermediate heat transport in other systems (SFR, LFR, VHTR) are investigated. Liquid salts offer two potential advantages: smaller equipment size, because of the higher volumetric heat capacity of the salts, and the absence of chemical exothermal reactions between the reactor, intermediate loop and power cycle coolants.

Liquid salt chemistry plays a major role in the viability demonstration, with such essential R&D issues as: the physico-chemical behaviour of coolant and fuel salts, including fission products and tritium; the compatibility of salts with structural materials for fuel and coolant circuits, as well as fuel processing material development; the on-site fuel processing; the maintenance, instrumentation and control of liquid salt chemistry (redox, purification, homogeneity), and; safety aspects, including interaction of liquid salts with various elements.

GIF progress in 2009

Significant progress was achieved in 2009. This included:

• Development of MSFR pre-conceptual designs and performance analysis of MSFR potential for starting with plutonium and minor actinides from PWR wastes (France).
• Laboratory scale processing of Ni-W-Cr alloys was recently demonstrated. The alloys were found to have acceptable workability and very good high temperature hardness (France, Auger et al., 2009). The whole potentialities of these kinds of materials as well as Hastelloy N have yet to be tested and characterized over the full range of temperatures and in presence of the fluoride salts.
• Corrosion tests of Ni-based alloys, (France, Fabre et al. 2009) and (ISTC, Ignatiev et al., 2008a).
• Better understanding of the PuF3 solubility in various carrier salts by means of thermochemical modeling (Euratom, Beneš et al., 2009).
• The material property database for molten and liquid salts was extended through experiments (Euratom) and theoretical calculations (Euratom, France). New experimental facilities were and continue to be developed (JRC-ITU).
• Significant improvement of fuel salt clean-up scheme (France).
• A code package for a fast MSR was developed (Hoogmoed, 2009) by coupling the 3-D time-dependent diffusion code DALTON with the thermo-hydraulics code HEAT (The Netherlands, TU-Delft).
• The optimal core configuration and salt composition of a moderated MSR that maximize the power density while keeping the self-breeding capabilities were found (The Netherlands, TU-Delft). New breeding gain definitions were developed (Nagy et al., 2010) that account for the unique behavior of the reactor. Some preliminary studies on the salt composition were published in (Nagy et al., 2008).
• Better understanding of the transmutation capabilities, dynamics and safety-related parameters, for fertile and fertile-free fuel concepts (IAEA, Ignatiev et al., 2008b).
• Demonstration of FHR performance and safety (USA).
• Construction of a fluoride salt test loop was initiated in the USA.
• An FHR component test plan was completed in the USA (Holcomb et al, 2009). The test plan provides a roadmap to the major technical demonstrations required to enable a test scale FHR to be built.
• Construction of a surrogate material compact integral effect test apparatus in support of a test scale FHR was initiated (USA). The new apparatus is intended to demonstrate the coupled thermal hydraulics response of FHRs to transients including loss of heat sink and loss of forced circulation.
• Criticality tests for the assessment of FHR fuel and core behavior (USA, Czech Republic).

A general discussion on these topics can be found in the Generation IV International Forum 2009 Annual Report (pages 52-59). More detailed explanations can be found in the bibliography below.  

Recent MSR research papers

Auger T., Cury R., Chevalier J.P. (2009), Development of Ni-W-Cr alloys for Gen IV Nuclear Reactor Applications, TMS annual meeting, 15-19 February 2009, San-Francisco, USA.

Beneš O., et al., (2008), Review Report on Liquid Salts for Various Applications, Deliverable D50, Assessment of Liquid Salts for Innovative Applications, ALISIA project, of the 7th Euratom Framework Programme.

Beneš O., Konings R.J.M., Actinide Burner Fuel: Potential compositions based on the thermodynamic evaluation of the MFX-PuF3 (M=Li, Na, K, Rb, Cs, La) system. J. Nucl. Mater. 377 (2008) 449.

Beneš O., Konings R.J.M., Thermodynamic evaluation of the LiF-NaF-BeF2-PuF3 system, J. Chem. Thermodyn., 41 (2009) 1086-1095.

Delpech S., et al., (2008a), Optimization of fuel reprocessing scheme for innovative molten salt reactor, paper presented at the October 2008 Molten Salts Joint Symposium, Kobe, Japan.

Delpech S., et al., (2008b), Actinides/Lanthanides Separation for the Thorium Molten Salt Reactor Fuel Treatment, paper presented at ATALANTE 2008, Montpellier, France.

Delpech S., et al., (2009a), Reactor Physics and Processing Scheme for Innovative Molten Salt Reactor System, J. of Fluorine Chemistry, 130, Issue 1, p. 11-17.

Delpech S., et al., (2009b), MSFR: Material issues and the Effect of Chemistry Control, 2009 GIF Symposium Paris France (2009).

Fabre S., et al., (2009), Corrosion of metallic materials for molten salt reactors, Proceedings of ICAPP’09, May 10-14 2009, Paper 9309, Tokyo, Japan.

Forsberg C.W., et al., (2007), Liquid Salt Applications and Molten Salt Reactors, presented at ICAPP, 13-18 May 2007, Nice, France.

Forsberg C.W., et al., (2008), Design Options for the Advanced High-Temperature Reactor, Paper presented at ICAPP, 8-12 June 2008, Anaheim, CA, United States.

Holcomb D.E., et al., (2009), An Analysis of Testing Requirements for Fluoride Salt-Cooled High Temperature Reactor Components, ORNL/TM-2009/297, November 2009.

Hoogmoed M.W., (2009), A Coupled Calculation Code System for the Thorium Molten Salt Rector, MSc. Thesis, PNR-131-2009-009, Delft, Netherlands.

Hron M., et al., (2008), Design Reactor Physical Program in the Frame of the MSR-SPHINX Transmuter Concept Development, paper presented at ICAPP, 8-12 June 2008, Anaheim, CA, United States.

Ignatiev V., et al., (2008a), Compatibility of selected Ni-based alloys in molten Li,Na,Be/F salts with PuF3 and tellurium additions, Nuclear Technology, Vol. 164, N°1, pp.130-142, October 2008.

Ignatiev V., et al., (2008b), Main Results of IAEA CRP on Studies of Advanced Options for Effective Incineration of Radioactive Waste: Case for Molten Salt Transmuter Systems, Paper presented at the 10th Information Exchange Meeting on Actinide and Fission Product Partitioning & Transmutation, 6-10 October 2008, Mito, Japan.

Mathieu L., et al., (2009), Possible Configurations for the TMSR and advantages of the Fast Non Moderated Version, Nuclear Science and Engineering 161, pp. 78-89.

Merle-Lucotte E., et al., (2009a), Minimizing the fissile inventory of the molten salt fast reactor, Proceedings of the International Conference Advances in Nuclear Fuel Management IV (ANFM IV), April 2009, Hilton Head Island, USA.

Merle-Lucotte E., et al., (2009b), Optimizing the Burning Efficiency and the Deployment Capacities of the Molten Salt Fast Reactor, Proceedings of the International Conference Global 2009 - The Nuclear Fuel Cycle: Sustainable Options & Industrial Perspectives, September 2009, Paris, France.

Nagy K., et al., (2008), Parametric studies on the fuel salt composition in thermal molten salt breeder reactors, Proceeding of PHYSOR 2008 International Conference, Interlaken, Switzerland, paper 277.

Nagy K., et al., (2010), Definition of breeding gain for molten salt reactors, Proceeding of PHYSOR 2010 International Conference, Pittsburg, USA, to be published.

Renault C., et al., (2009), The Molten Salt Reactor (MSR) in Generation IV - Overview and Perspectives, 2009 GIF Symposium Paris, France (2009).

Salanne M., et al., (2009), Transport in molten LiF-NaF-ZrF4 mixtures: a combined computational and experimental approach, Journal of Fluorine Chemistry, 130, pp. 61-66.

Zherebtsov A., et al., (2008), Experimental Study of Molten Salt Technology for Safe, Low-Waste and Proliferation Resistant Treatment of RadioactiveWaste and Plutonium in Accelerator Driven and Critical Systems, ISTC-1606 Project, Final Report, International Scientific Centre, Moscow, Russian Federation.

E-mail contact: msr@gen-4.org

Back to top