Development of improved ferritic/martensitic creep-resistant steels for nuclear energy

Puype Athina

Promoter

Petrov Roumen, (UGent), roumen.petrov@ugent.be

SCK•CEN Mentor

Malerba Lorenzo
lorenzo.malerba@sckcen.be
+32 14 33 30 90

SCK•CEN Co-mentor

Bonny Giovanni
giovanni.bonny@sckcen.be
+32 14 33 31 98

Expert group

Structural Materials Modelling and Microstructure

PhD started

2014-09-01

Short project description

Ferritic/martensitic (F/M) steels have been proposed as fuel assembly materials at the time of the fast reactor programmes in the 1970s-1980s because, by swelling significantly less than austenitic steels under irradiation, they would allow the fuel to be kept for longer time in the reactor, thereby increasing the burnup[1]. Burnup increase has a number of quite obvious advantages: more efficient use of the fuel (more energy produced from the same fuel pins), consumption of not only uranium and plutonium, but also minor actinides, with subsequently reduced amount of waste (in a closed fuel cycle scenario …), etc. In addition to superior resistance to irradiation, F/M steels have also better thermal properties, in particular higher thermal conductivity than austenitic, thereby offering better guarantees in the case of thermal transient, especially an accidental one.

The current goals of increasing both thermodynamic efficiency and safety in GenIV reactors require that fuel assembly materials should have good elevated-temperature mechanical properties (so, good creep resistance), while stably offering good fracture toughness at all temperatures, including under irradiation. Fusion reactors have in this respect very similar requirements, in addition to the stringent limitation of reduced activation. The loss of creep resistance above 550°C in existing F/M steels, used or usable in nuclear energy, is one factor that limits the operating temperature of fast reactors. Austenitic steels, at least 316L, allow in this respect somewhat higher temperature[2], even though the difference is not spectacular. Thus, F/M steels with superior creep resistance would certainly find application for the supporting component of the core of fast reactors, as well as for fusion applications, given that their resistance to irradiation up to high dose, swelling in particular, but also irradiation creep, is higher. Already now, standard F/M steels such as EM10 are planned to be used for in-core supporting components of the future French sodium fast reactor, ASTRID, such as wrappers.

In order to increase the creep resistance and therefore increase the operating temperature, oxide dispersion strengthening (ODS) of F/M steels is currently being considered. However, ODS steels can currently only be produced from powder technology by mechanical alloying followed by extrusion. This limits the industrial scalability of their production. Moreover, to date it remains challenging to guarantee the uniform distribution of nanometric dispersoids that would ideally be required, as well as to remove the anisotropy of the properties of these alloys. It is also difficult to guarantee reasonable uniformity of properties from heat to heat. Finally, fracture toughness remains quite severely penalized even at high temperature: it is therefore unlikely that these alloys will ever be usable for any other reactor component except cladding. Thus, ODS steels remain a possible alternative as nuclear reactor core materials, but projected far in the future as a long-term perspective, and most likely their use will not be considered for structural components.

In recent times an intermediate solution between current F/M steels and ODS steels has been considered, which would remove the problems of fabricability. The idea is to push up the upper operating temperature limit of conventionally produced F/M steels, by improving their thermal creep-resistance, while keeping their composition compatible with the requirement of nuclear energy, including as much as possible low activation. This research line has been pursued in the US by R. Klueh and co-workers at ORNL, is being studied at CEA in France, where a few PhD theses are devoted to this route, and has been included as one of the priorities of the current European fusion roadmap. In no case, however, for the moment, has any steel-maker been involved in such type of research. Commercial creep-resistant steels of superior capability in terms of creep-resistance and therefore with higher operation temperature limits do exist, in fact (so called 4th Generation creep-resistant steels), for example P92, HCM12A, and more recently NF12 and SAVE12. However, the composition of these steels is clearly incompatible with their use in nuclear energy, because of activation and in general neutronic reasons (e.g. the use of boron or cobalt), but also because of the presence of elements that lead to pernicious radiation effects in terms of embrittlement (nickel, copper, …).

It is believed that 4th generation steels of superior creep-resistance and compatible with nuclear energy use can be produced by tuning the composition and applying specifically tailored thermal-mechanical treatments, while considering advanced steel production routes currently used for industrial applications but unknown to the nuclear sector.


[1] Amount of fuel that is actually 'burnt', i.e. used, which is currently limited to 20%, the limiting factor being not the fuel itself, but the material supporting it, i.e. the cladding.

[2] Ni-based alloys are not advisable as fuel assembly materials for neutronic reasons, and even less can they be considered for fusion application, given the high activation and He production by transmutation of Ni. In this respect, also austenitic steels are not of use for fusion applications.

 

Objective

The goal of the present proposal is thus to revisit the composition of creep-resistant steels, avoiding "nuclear-sensitive" elements, while elaborating specific production routes aimed at producing a microstructure that guarantees high creep-resistance as well as stability under irradiation.

The choice of the elements to be added or avoided should be made based on several criteria:

  • Limit the level of activation under irradiation and of He production (so, compositions such as the one of Eurofer or F82H can be taken as a starting point);
  • Avoid those elements that are known, from practical experience and from existing advanced atomistic models, to interact strongly with radiation-defects in such a way to potentially exacerbate low temperature embrittlement;
  • Consider elements that, based on both metallurgical experience and advanced atomistic models, seem a priori to offer potentially beneficial effects in terms of solution strengthening, stability of carbonitrides, but also removal or retention of radiation defects, with a view to reducing low temperature embrittlement, as well as swelling.

Techniques of industrial use aimed at keeping under control the grain size will also be considered, with a view to increasing the capability of absorbing radiation defects and increasing the radiation resistance of the steel.