XII MEETING ON "REACTOR PHYSICS CALCULATIONS IN THE NORDIC COUNTRIES", MAY 2005, 1/8 Managing Xenon Oscillations in Monte-Carlo Burnup Simulations of Thermal Reactors Jan Dufek* and Waclaw Gudowski Department of Nuclear and Reactor Physics Royal Institute of Technology (KTH) Roslagstullsbacken 21, S-10691, Stockholm, Sweden *jan.dufek@neutron.kth.se Abstract The paper summarizes our research in the stability of the Monte Carlo burnup calculations of nuclear reactor cores. Calculations with the Monte Carlo burnup codes MCB and Monteburns show that the codes have a stability problem, if the core model allows various fuel burnup at various positions, and the time steps are longer than around one day. The instability is caused partially by a spatial oscillation of xenon mass distribution, and partially by a spatial oscillation of burnup distribution. The paper discusses the instability in detail, and proposes a solution. Particularly, the Monte Carlo calculation of the steady-state core conditions is described. All calculations have been made for thermal reactors. 1. Introduction Problems of burnup calculation of reactor fuel cycles are discussed in the paper. We particularly focus on the Monte Carlo approach which is very suitable for reactor research area due to its ability to use “continuous” cross-section data. In the recent years, several new Monte Carlo burnup codes have appeared; e.g. MCB [1] or Monteburns [2] which represent two different approaches. MCB code developed at KTH in collaboration with the UMM in Krakow, is a Monte Carlo Continuous Energy Burnup Code for a general-purpose use to calculate a nuclide density time evolution with burnup or decay. MCB integrates the well-known code MCNP [3], which is used for neutron transport calculations, and a novel Transmutation Trajectory Analysis (TTA), which serves for density evolution (burnup) calculation. MCB collects the information for the burnup calculation during the particle transport simulation, and therefore uses continuous cross-section tables from MCNP. Monteburns links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2 [4]. The principle function of Monteburns is to transfer one-group cross-section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. Although the codes vary a lot, the basic principle is the same for both of them. They divide the burnup calculation into a number of time steps. At each time step the neutron transport is simulated in the core, and the fuel composition is adjusted according to the obtained information which includes the neutron flux and neutron cross-sections. Unfortunately, the Monte Carlo approach has a big disadvantage contrary to the deterministic approach – it requires a considerable processor time to simulate the particle transport with satisfactory precision. Thus, one has to specify rather small amount of time steps which might be relatively large