Studies on the assessment and validation of reactor dynamics models used in Finland: Dissertation

Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes, TRAB and SMATRA, have been examined from two points of view: First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In this study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. It appeared that the models of the pressure loss terms in the codes predict well the observed behaviour in a VVER specific test bundle. Well-known critical heat flux correlations have been tested against four data sets, and one of the correlations was modified to adapt the critical heat flux table presented by Kirillov. Unlike most other critical heat flux correlations, the models of Gidropress and, in particular, Smolin proved to cover a very wide parameter range with predictions mostly on the conservative side. These two correlations are used in the studied codes. On the other hand, some uncertainty must be associated with the modelling of the post dryout heat transfer, although the model used in the codes is probably on the conservative side in the VVER rod bundle. The SMATRA code reliably reproduced the four transients chosen as validation cases for the Loviisa VVER-440 nuclear power plant. The minor deviations observed in the simulations may, at least partly, be associated with uncertainties in the plant data. In the quoted safety analyses the axially one-dimensional neutron kinetics model of the codes, appended with the transverse synthesis model, appeared to be valuable in the calculation of the reactivity initiated accidents and the anticipated transients without scram. Some boron dilution cases, which may be critical for the safety of the power plant, have been singled out for particular investigation in this study.