Uncertainty Quantification for the Doppler Reactivity Feedback Coefficient of MYRRHA

Safety analyses of MYRRHA require calculating the reactor temperature feedback coefficients of reactivity associated with power operation. Within this framework, a significant emphasis is put on the quantification of the reactivity coefficient uncertainty, which takes into account different uncertainty sources. This work reports the investigation of the fuel Doppler coefficient of reactivity of MYRRHA implemented with the Serpent 2 Monte Carlo particle transport code. Nuclear data and the MYRRHA mixed-oxide (MOX) fuel composition are selected as major sources of uncertainties, and their impact on the Doppler coefficient is assessed with a stochastic sampling approach. Nuclear data covariance matrices are taken from the JEFF-3.3 evaluated library and propagated with the SANDY code.