Speaker
Stuart Maloy
(LANL)
Description
The Fuel Cycle Research and Development program is
investigating methods of dealing with transuranics in
various fuel cycle options and is supporting the development
of next generation Light Water Reactor (LWR) fuels. To
achieve this goal, new fuels and cladding materials must be
developed and tested to high burnup levels (e.g. >20%) and
under accident conditions. To achieve such high burnup
levels the fast reactor core materials (cladding and duct)
must be able to withstand very high doses (greater than 200
dpa) while in contact with the coolant and the fuel. Thus,
these materials must withstand radiation effects that
promote low temperature embrittlement, radiation induced
segregation, high temperature helium embrittlement,
swelling, irradiation creep, corrosion with the coolant, and
chemical interaction with the fuel (FCCI).
To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-700°C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510°C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings and liners are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions. Recent progress in high dose irradiated materials testing and advanced radiation resistant materials development will be presented.
To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-700°C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510°C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings and liners are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions. Recent progress in high dose irradiated materials testing and advanced radiation resistant materials development will be presented.
