Neutron and Gamma Fluence and Radiation Damage Parameters of Ex-core Components of Russian and German Light Water Reactors

Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. This work aims to facilitate for some reactor types the evaluation of the importance of gamma radiation for embrittlement studies. Absolute neutron and gamma fluence spectra had been calculated for two core loading variants of the Russian PWR type VVER-1000, for a German PWR of Konvoi-Type and for a German BWR. Based on the calculated spectra several fluence integrals and radiation damage parameters were derived for the region of the midplane azimuthal flux maximum for different radial positions between the core and the biological shield, particularly, the inner and outer PV walls, the ¼ PV thickness and the surveillance positions. The relative contributions of gamma radiation to the sums of gamma and neutron contributions are of special interest. As damage parameters the displacements per atom of iron are given separately for neutrons and gammas as well as some rough estimations of the numbers of freely migrating defects. To get some notion about the uncertainty of the obtained dpa, the calculations were performed using different dpa cross section evaluations. Additionally, gamma produced dpa were calculated by means of the Monte Carlo code EGS.
Another parameter of practical interest for pressure vessel dosimetry, the contribution of photofission to the total number fissions, was calculated for the detector reactions 237Np(n,f) and 238U(n,f).
Most of the calculations were performed using a 3D synthesis of 2D/1D-flux distributions obtained by the DORT code with the BUGLE-96 library. To increase the reliability of the evaluations some of the calculations were repeated by different laboratories. For two reactors the influence of the group approximation on the calculation results was investigated by parallel calculations with the continuous energy Monte Carlo code MCNP using nuclear data from the library ENDF/B-VI.
The results were compared and the reasons of found discrepancies were discussed.