Jean Couturier, and 12 co-authors conclude that in terms of the safety demonstration of Generation IV (Gen IV) Reactors, the available analyses produced by the designers mean that, given a low-power density in the core, it is possible to envisage residual heat removal by radiative heat transfer from the reactor core and a very low risk of fuel cladding damage. Furthermore, fuel meltdown as such can be excluded, which is a favorable aspect of V/HTRs. However, this observation would need to be supplemented by a comprehensive assessment of the risk of reaching core temperatures above 1,600°C, which could lead to extensive damage to the fuel particle coatings, taking into account the risks of total cooling system failure, including the reactor pit cooling system (which may operate passively in some designs).
The dosimetry consequences of deposition in the primary system of carbon dust from the core, which could contain radioactive elements, need to be assessed, in particular with regard to in-service maintenance (including in-service inspection). Finally, with reference to the events that occurred on the Fukushima- Daiichi nuclear power plant, the risk of air or water ingress into the reactor, in the event of a LOCA or flooding, would need to be covered.
Gas-cooled Fast Reactors (GFRs)
GFRs do not benefit from any experience feedback (as no power or experimental GFR has ever been built or operated). High temperatures (over 800°C at the core outlet) are targeted in order to obtain high power-generation efficiencies. However, there is currently no fuel that is compatible with such conditions, which constitutes a technological obstacle. R&D work is underway.Compared with SFRs, favorable aspects include the use of an inert gas (helium) for core cooling, a much weaker "void effect", and the general absence of materials that may react violently with air or water. However, in terms of the safety demonstration, prevention of core melt is a major subject for GFRs. In particular, managing the consequences of a LOCA is a major safety issue for the design, which could lead to complex architecture for protection and safeguard systems, necessarily involving active cooling systems. Furthermore, these systems would have to be highly robust with respect to events such as those that occurred on the Fukushima-Daiichi nuclear power plant. In addition, water ingress into a GFR could lead to a reactivity accident and appropriate prevention would be required.
Lead-or Lead/Bismuth-Cooled Fast Reactors (LFR).
For LFRs, relatively limited experience feedback has been acquired,uniquely for the reactors on Soviet Alfa-class submarines. Three of these reactors suffered serious damage, including one case of partial core meltdown. LFRs use molten lead or a lead-bismuth eutectic alloy (called LBE) as coolant, which have the advantage of not reacting violently with air or water. There do not seem to be technological obstacles to overcome for LFRs. Management of the risk of corrosion by lead or LBE is nonetheless a key safety issue, but a procedure exists to manage this risk, based on maintaining an appropriate oxygen concentration in the lead and then purifying the lead to remove any oxides formed. This procedure seems to have been validated during operations of the submarine reactors mentioned above, although its suitability for the operation of a large reactor would need to be confirmed. It should also be stressed that the oxygen concentration in lead or LBE would need to be carefully adjusted, as the presence of oxygen (or hydrogen) in lead or LBE increases their viscosity,which can cause overheating in the core.
Special attention would need to be paid to preventing the lead or LBE from freezing on an LFR. In addition, the risk of embrittlement of metal structures by lead or LBE and their sudden failure is a major issue that has been the subject of research.In terms of the safety demonstration, the concerns mentioned above for SFRs also seem to apply to LFRs, albeit with the favorable aspect of a lower power density in the core than for an SFR (approximately three times less). With regard to the events which occurred on the Fukushima-Daiichi nuclear power plant, although lead has the advantage of not reacting with water, it is toxic and robust design provisions would need to be taken to prevent releases of lead into the environment,for example in the event of a break in a loop containing lead or LBE in a flooded room. Ultimately, the operation of an LFR seems complex, with the need to manage the risk of reactor overheating due to excessive oxygen or hydrogen (increasing the viscosity of lead or LBE) combined with the need to manage corrosion risks by using a technique based on injecting oxygen into the lead or LBE. While this is conceivable for an "overmanaged" operational context or an R&D context (MYRRHA project), industrial operation could prove more problematic.
Molten Salt Reactors (MSR)
Experience feedback on MSRs is limited to two experimental reactors built and operated in the USA in the 1960s: the Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment(MSRE), which operated for 13,000 hours.The CNRS is currently developing the Molten Salt Fast Reactor (MSFR) project, using uranium-233 and thorium-232 dissolved in a lithium-based fluoride salt. At the current state of the project, the MSFR seems to benefit from favorable neutronics characteristics. Reactor shutdown would be provided by fuel salt drainage and passive residual heat removal from the dump tanks is being examined. There does not seem to be a technological obstacle, even though corrosion by the salt is a significant issue that has been the subject of R&D since the 1950s. There is a procedure for managing this risk, which seems to have been validated during operations on the American MSRE reactor. Clearly, this procedure would need to be validated for larger reactors. Alloys that are compatible with the use of salts at the temperatures targeted for MSRs (700°C) are available. However, given the originality of the MSR design as described by the MSFR project, associated with the use of a fluid that is both fuel and coolant, there would need to be a revision of defense-in-depth and the safety approaches that have been produced and implemented for current reactors. Special attention would need to be paid to characterization of the radioactive contents (in terms of aspects such as the locations, quantities and spectra of radionuclides) in the system considered as a whole, including the reactor itself and the reprocessing unit, which will contain fuel salt and fission products.
This analysis should systematically substantiate the architecture and design of the barriers and identify risks of criticality, in particular in the event of flooding for example. Finally, use of beryllium salts would be discouraged, given the toxicity of beryllium and its characteristics in terms of neutron moderation (and the consequent risk of criticality).
Super Critical-Water-cooled Reactors (SCWR)
While super critical water technology has been used in conventional industry, no SCWR has ever been built and operated. The SCWR concept has some very economically attractive characteristics, mainly as it can be considered as an innovative version of light-water reactors, in particular BWRs. SCWRs reuse a certain number of the general characteristics of these reactors (such as control of reactivity using absorber rods and emergency borated water, and a safety injection system). However, a certain number of subjects merit specific R&D before a position can be taken on its industrial viability. Furthermore, the engineering associated with this design appears relatively complex and current research is focusing primarily on subjects that deserve development and is also aiming to simplify the design. In principle, the SCWR should also be able to integrate the main safety principles that are being developed on third-generation light water reactors, and experience feedback from the accident that occurred on the Fukushima-Daiichi nuclear power plant, in particular the improvements that may be made to BWRs.
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