Selection of materials and main issues
For the selection of candidate materials for MYRRHA, the option has been taken to consider industrially available and qualified materials, rather than to develop and optimise innovative materials. Since at relatively high temperatures (> 500-550 °C) no proven available materials under relevant irradiation conditions exist today, so operating temperatures for the first demo concepts are to be limited (up to 480 °C for the lead fast reactor prototype).
The materials qualification programme for MYRRHA shows commonalities with the corresponding work for the GEN IV Lead Fast Reactor. The structural materials issues to be considered are related to design and safety requirements, where gaps and open questions concerning materials performance under representative conditions (e.g. high temperature, high burn-ups, and aggressive environment) are identified. These issues can be classified into five areas:
- Workability and fabricability of materials and components
- Mitigation of corrosion by liquid metal
- Embrittlement by interaction with the environment
- Irradiation effects (embrittlement, creep, swelling, …)
- Transferability of the experimental results to the actual MYRRHA machine
Based on literature data on mechanical and thermal properties, irradiation performance, fabricability and availability, high Cr ferritic/martensitic steels and nuclear grade austenitic steels have been selected for evaluation.
The large available database on mechanical properties and irradiation effects suggests austenitic steels, especially low-carbon grade, as candidates for components operating at relatively low temperatures and low irradiation fluence, e.g. the reactor vessel.
FMS (T91) appears to be among the best candidate materials for fuel cladding and highly irradiated structures, because of their resistance against swelling and creep under high fast neutron flux. However, to respect the planning of MYRRHA, it is possible that for the first core loadings qualified austenitic materials (such as the 15-15Ti stabilised steels) are chosen for the fuel cladding. To have a qualified T91 material for internal structures and especially for fuel cladding in the mid-term, a thorough demonstration and qualification programme should be pursued.
As the risk involved in the qualification of a new material, especially for fuel cladding, is high, a fall-back option is taken by selecting a qualified cladding material from the fast breeder reactor programmes. In the latter case, only the limits imposed by interaction with the innovative coolant environment in MYRRHA has to be ascertained, as the material's performance as fuel cladding has been qualified in representative conditions of sodium cooled fast reactors (up to accumulated peak dose ~100 dpa).
Materials qualification strategy for MYRRHA
The materials qualification strategy for MYRRHA relies on the strong common trunk with the GEN IV LFR research and development programmes, complemented with targeted actions towards specific needs in the MYRRHA design. Within the five areas identified above, the following actions are taken:
Workability and fabricability: common trunk on welding-joining techniques and cladding fabrication in EC FP6-FP7 projects; MYRRHA component specific actions will consist in analysis of prototypes from pumps and heat exchangers before and after testing.
Corrosion by liquid metal: common trunk on corrosion and corrosion protection in LBE with FP6 and FP7 projects; a specific action is to be set-up to integrate the available data on corrosion into a descriptive model relevant for the corrosion issues in the coolant circuit and to propose mitigation measures, including their validation. Another specific action is to ascertain the potential corrosion effects on the secondary side of the heat exchangers by general corrosion and stress corrosion experiments.
Embrittlement by interaction with the environment (unirradiated materials): common trunk with FP6 and FP7 projects on ADS and LFR. A specific action is to develop and validate experimental tools to measure fracture toughness in LBE environment, with special attention to weld materials.
Irradiation effects: completion of the database on irradiated materials, mainly from fast reactor programmes, with targeted experiments (especially on T91) in LFR relevant conditions. This is partially covered by the FP6 and FP7 projects ongoing; a dedicated action will be taken in order to complement this database with data on the combined effect of irradiation and exposure to LBE at the lower temperature range (300-350 °C) on the mechanical behaviour (especially fracture toughness) of the material. This requires irradiation of materials in LBE up to dose levels where the degradation of fracture toughness is expected to saturate (by the saturation of the irradiation induced hardening), i.e. between 5-10 dpa (representative maximum dose on core support plate).
Transferability of experimental results: the development of physics based models for extrapolation of results on the effect of irradiation on microstructure and mechanical behaviour of the Fe-Cr steels is fully embedded within European collaborative projects. On the long-term, dedicated actions to validate MYRRHA relevant prediction of the multi-scale models will be undertaken. Once validated, the modelling effort will support the design of the MYRRHA by evaluation of the expected effect of modification of the operational conditions on the material property evolution (power upgrades, operation temperature variations, transition from ADS to fast fission spectrum…).