Advanced Neutron Dosimetry on VVER-440 Ained to Reactor Equipment Load Evaluation During Life Time Prolongation

Lifetime of non-restorable reactor equipment (RPV, in- and ex-vessel constructions, reactor support structures) of VVER-440 of first generation may restrict plant-life at whole. Concerning VVER, service life extension has touched the first generation of VVER-440 type reactors. Nowadays, all Russian first generation VVER-440 operated with reduced core by using dummy assemblies except reactor of Unit 4 of Novovoronezh NPP, wich are under operation more than 45 years. In comparison with other power units, the full-core loading scheme of this unit results in the highest neutron fluence accumulated over whole operation period. The analysis of degradation of material property of the equipment under irradiation are urgent in aspect of confirmation of safety of operating VVER units of similar type, and also in connection with an opportunity of further life-time prolongation. Another important aspect of dosimetry research on the Unit 4 of Novovoronezh NPP reactor is the verification and validation of calculational methods used for the prediction of the maximum neutron fluence on the VVER-440 pressure vessel.
Ex-vessel neutron dosimetry at NPP with VVER is used for a purpose of validation of calculations 3D distribution of neutron fluence on the reactor equipment. The reliability of such type validation depends on a quality of measured data and a comprehensive analysis of calculational model. The ex-vessel measurements permit to evaluate real conservatism of RPV neutron fluence evaluation, and, if possible, reduce conservative safety margins. For this purpose special advanced neutron dosimetry research were implemented at Unit 4 of Novovoronezh NPP (NV-4). Neutron activation measurement were performed in ex-vessel positions for different fuel cycles. The measurements of detectors activities also have been performed by different participants.
The purpose of the research was to perform an intercomparison of independent calculational results and comparison with reference data of the full-scale VVER-440 ex-vessel experiment. The paper deals with an analysis of a comparison of results of independent calculations by deterministic and Monte-Carlo codes between themselves and with experimental results. New and improved results of calculations analysed in the paper were run with codes and applied libraries DORT/BUGLE-96, TRAMO/ENDF/B-VII, MCNP/ENDF/B-VII, TORT/BUGLE-96, KATRIN/BGL-440, KATRIN/V7-200N47G. Calculated pure 3D or 3D synthesis multigroup neutron fluence rates in positions of experimental capsules provided by every calculator were used as basic calculational information. Based on the calculated fluence spectra the various reaction rates were computed using the dosimetry cross sections from the library IRDF-2002. The results of intercomparison of different calculation procedures and different experimental techniques give allowance to estimate of the reliable calculational and experimental uncertainties.
This verification of calculated methods provides the reliability of assessment of maximum neutron fluence on reactor equipment accumulated over whole service life period. In addition, with regard to a lifetime extension of Russian VVER-440 type reactors, advices for optimal core loading schemes will be derived. To approve the reliability and/or conservatism of calculation procedures the improved uncertainty-accounted conservative evaluation of RPV radiation loading parameters has been performed. This approach takes into account over- or underestimated C/E ratio and uncertainties of measured and calculated results. Based on such approach the analysis of end-of life-time fluence estimation has been demonstrated in the paper. The paper describes the features of the calculations, compares their results with those of the neutron activation measurements and presents neutron fluence values that were computed.