Reprocessing and Dissolution
ASGARD is scheduled to look at oxide, nitride and carbide nuclear fuels. The breakdown in these technical domains forms the backbone of the project.
The oxide dissolution and separation strategy is a fairly mature process being dealt
with and optimised in the FP7 EURATOM ACSEPT project. New separation strategies have been tested on genuine spent fuel and the selected processes will be evaluated for industrial implementation. Whereas the above is valid for actinide oxide fuels, such as MOX and/or Minor Actinide containing MOX, the dissolution and separation issues for inert matrix fuels containing ceramic MgO or metallic molybdenum (Mo), has not been investigated coherently. In order to be complementary to ACSEPT, the ASGARD project focuses therefore mainly on the Inert Matrix Fuels (IMF) with molybdenum or magnesium-oxide. It is of crucial importance to take into account the behaviour of the matrix elements in the dissolution and separation processes and to check their compatibility with the future immobilisation (impact on the stability of the waste and amount of generated waste). In addition, the use of molybdenum as an inert matrix poses additional challenges with respect to its redox chemistry, the need to avoid precipitation or co-precipitation, and the necessity to recover the isotopically tailored Mo material.
For the assessment of inert matrix fuels, some basic studies will be performed to assess the dissolution of minor-actinide containing oxides, specifically with high americium and plutonium content.
Nitride fuels constitute a high performance alternative to oxide fuel. Major advantages include a higher actinide density and a combination of high thermal conductivity with high melting temperature. The latter are particularly important in the context of transmutation in Generation IV reactors, since the addition of minor actinides to the fuel is detrimental for reactivity feedbacks. The resulting increase in fuel temperature under power transients is more easily accommodated by the larger margin to failure of the nitride fuel. The solubility of actinide nitrides in nitric acid is good in general. A major issue that needs to be addressed is how N-15 recycling is to be implemented.
Carbide fuels have been considered as fuels for Fast Reactors since the 1960s and many irradiation experiments on U and U/Pu carbide have been performed globally, with recent successes particularly for the Fast Breeder Test Reactor at Kalpakkam, India. The physical properties of carbide fuel make them attractive because they are conducive to high specific rod powers with relatively low fuel centre temperatures: start of life power capability is increased, power-to-melt margin is increased
and fatter (more economic) pins are facilitated. Because fuel temperatures are low, the fuel suffers little or no restructuring and the release of fission gases and volatile fission products is low. Under transient conditions, high thermal conductivity and low thermal expansion offer certain advantages i.e. lower fuel/clad mechanical interaction (FCMI) levels in over power transients and lower cladding temperatures in under cooling events. Operationally, better internal breeding reduces reactivity
decline over the fuel lifetime which eases control rod management. Carbide fuel chemical properties provide the advantage that under normal operating conditions
fuel/stainless steel clad chemical interaction is virtually absent; additionally, swelling due to fuel/coolant interaction in a failed pin should not occur to the benefit of minimising contamination of the reactor circuit. Thus two sources of concern present with oxide pins are potentially eliminated.On the down side there are a number of evident problems with carbide fuel. Performance limitations are of concern because of the significant potential for unacceptable fuel/clad mechanical interaction (FCMI) due to the high swelling propensity from gases trapped within the fuel matrix
leading to the possibility of extensive reaction between fuel and cladding at higher transient temperatures, thus limiting the carbide fuel centreline pellet temperature and the overall economic performance of the reactor. It is then clear that if the fuel swelling mechanism can be avoided/reduced carbide fuel could have significant advantages towards FCMI compared to oxide fuel.
In addition, we have an extensive education and training programme to give the possibility for young researchers and professionals to learn and train in these highly specialized nuclear applications. This is especially important since the “nuclear population” is aging and replacement is scarce.