Testing of multigroup neutron cross-section libraries on neutron transmission through the VVER-440 vessel

The verification of calculated neutron fluence onto the VVER-440/230 pressure vessel is a very timely task, especially considering that some of these reactors have been operating for the major part of the reactor design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux, the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results are presented for Mn-54-induced activity of scraps from the inner wall of the Unit I reactor pressure vessel after the 18th cycle and from detectors irradiated behind the vessel during the 18th cycle of Unit I at Kozloduy Nuclear Fowler Plant as well as activity attenuation through the VVER-440/230 pressure vessel. Neutron cross-section libraries generated on the basis of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and the attenuation coefficient demonstrates that the neutron fluence calculations by the libraries based on ENDF/B-VI are more reliable than ones based on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the VVER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the VVER-440 vessel with dummy cassette loading.