A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.

Molten salts (fluorides or chlorides) have been taken in consideration very soon in nuclear energy production researches. This was initially due to their advantageous physical properties: good heat transfer capacity, radiation insensitivity, high boiling point, .. and they can be used in various situations: heat transfer, core coolants with solid fuels, liquid fuel in molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing, fusion. Molten salt reactors which are one of the six innovative concepts chosen by the Generation IV international forum may be particularly interesting in the case of waste incinerators or of the thorium cycle. As the neutron balance is very tight, the possibility to quickly extract poisoning fission products is very attractive. The most important questions addressed to demonstrate the scientific feasibility of Molten Salt Reactor will be reviewed.

In the last 5 years, there has been a rapid increase in interest in the use of hightemperature (700 to 1000°C) molten and liquid fluoride salts as coolants and for other functions in nuclear systems. This interest is a consequence of new applications for high-temperature heat and the development of new reactor concepts. These salts have melting points between 315 and 565ºC; thus, they are of use only in high-temperature systems. Nitrate salts with a peak operating temperature of ~600°C are the highest-temperature commercial liquid coolant available today; therefore, the development of higher-temperature salts as coolants offers the possibility of new nuclear and nonnuclear applications. These salts are being considered for intermediate heat-transport loops between all types of high-temperature reactors (helium and salt cooled) and hydrogen production systems, oil refineries, and shale oil processing facilities. Historically, steam cycles with temperature limits of ~550°C have been the only efficient method to convert heat to electricity. This limitation produced few incentives to develop high-temperature reactors for electricity production. However, recent advances in Brayton gas-turbine technology now make it possible to convert higher-temperature heat efficiently into electricity and thus have created the enabling technology for more efficient cost-effective hightemperature reactors. The near-term Advanced High-Temperature Reactor uses a graphite-matrix, coated-particle fuel and a liquid salt coolant. A longer-term potential exists for a liquid-salt-cooled fast reactor that uses metal-clad fuel and a liquid salt coolant. The molten salt reactor (MSR), with the fuel dissolved in the molten salt coolant, is receiving attention because of (1) the advancing salt-coolant technology and Brayton cycles that improve the economics; (2) advances in salt chemistry that enable the development of fast-spectrum MSRs with the safety advantages of large negative void coefficients; and (3) the interest in actinide burning where MSRs avoid the need to fabricate fuel of highly active actinides. Last, there is a developing interest in liquid-wall fusion machines with much higher power densities than solid-wall fusion machines.

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In recent years great interest has been displayed, worldwide, for Accelerator Driven Sub critical reactors (ADS) to incinerate the minor actinides generated by the existing energy producing reactors. In sub critical systems, the effective neutron multiplication factor is lower than 1.0 and the neutrons otherwise required to maintain the chain reaction, can be put to other uses, in particular, the destruction of nuclear wastes such as minor actinides (MA). One of the major advantages of such ADS systems is that it can be operated with very high M.A content without jeopardizing the overall safety due to a small effective delayed neutron fraction, a small Doppler temperature coefficient and possibly also a large void coefficient depending on the chosen coolant. This enhanced safety however prerequisites at all time a sufficient subcriticality margin. Reliable reactivity monitoring techniques are hence required to achieve this goal. The MUSE-4 program is a series of low power experiments carried out at the CEA-Cadarache MASURCA facility to investigate the various methods leading to the measurement of the reactivity level and associated kinetic parameters such as the effective delayed neutron fraction. The aim of this paper is to present the results obtained with a method which directly gives the ratio, for a sub critical assembly, between the reactivity ρ and the effective delayed neutron fraction β eff . By combining these results to those obtained with the k p -method for the prompt neutron multiplication coefficient, we have access to the parameters which govern the prompt and the slow kinetics of a sub critical assembly. These parameters can be obtained without reference to any calibration measurement in critical configuration. It opens the way to the control of larger sub critical demonstrators which are operating with fuels which cannot be used in critical reactor, and, thanks to sub criticality, which are characterized by a deterministic safety.