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The problem of ensuring the integrity of VVER type reactor equipment is now most significant in justifying the safety of nuclear power plant units and the extension of their service lifetime to 60 years and more. This issue primarily concerns old NPP power units with VVER-440s and VVER-1000s. The justification of VVER equipment integrity depends on the reliability of estimating the degree of equipment damage. One of the mandatory requirements, providing the reliability of such estimation and also the evaluation of VVER equipment lifetime, is the monitoring of equipment radiation loading parameters. Relative to this requirement is the challenge of justifying the normative parameters used for estimating reactor pressure vessel metal embrittlement as the fluence and fluence rate of fast neutrons with energies above 0.5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by the Scientific and Engineering Centre for Nuclear and Radiation Safety for all Russian NPP. As a result of this activity, the Scientific and Engineering Centre for Nuclear and Radiation Safety has recently elaborated on new approaches aimed at monitoring the radiation load of all Russian VVER equipment. This paper describes these approaches and explains the means of their implementation during monitoring procedures.

The lifetime of nonrestorable reactor equipment (e.g., reactor pressure vessels, in- and ex-vessel constructions, reactor support structures) of a first-generation water-water energetic reactor, VVER-440, may restrict plant life as a whole. Service life extension has affected the first generation of VVER-440 type reactors. Today, all Russian first-generation VVER-440s that have been in operation more than 45 years run with reduced cores by using dummy assemblies, except reactor Unit 4 of the Novovoronezh nuclear power plant. In comparison with other power units, the full-core loading scheme of this unit results in the highest neutron fluence accumulated over the whole operation period. Another important aspect of dosimetry research on Unit 4 of the Novovoronezh nuclear power plant reactor is the verification and validation of calculational methods used for the prediction of the maximum neutron fluence on the VVER-440 pressure vessel.

The problem of ensuring the integrity of VVER type reactor equipment integrity is now most significant in connection with justifying the safety of the NPP units and the extension of their service life-time to 60 years and more. This issue primarily first of all concerns long term operated NPP power units with VVER-440s and VVER-1000s. The justification of the VVER equipment integrity depends on the reliability of estimation of the degree of the equipment damage. One of the mandatory requirements [1], providing the reliability of such estimation, and also the evaluation of VVER equipment life-time, is the monitoring of equipment radiation loading parameters. Relative to this requirement there is a problem the challenge of justification of such the normative parameters, used for an estimating of the reactor pressure vessel (RPV) metal embrittlement, as the fluence and fluence rate of fast neutrons with energies above 0,5 MeV. Compliance with these requirements is analyzed during regular monitoring of radiation load parameters, which is performed by SEC NRS for all Russian NPP from the regulatory point of view. As a result of this activity, SEC NRS has recently elaborated one of the new approaches aimed to monitoring the radiation load of all equipment of Russian VVERs. The paper describes these approaches and shows the way of their implementation during monitoring procedures.

Prediction of radiation load parameters (fluence, fluence rate, neutron energy spectrum) of reactor components (e.g., reactor pressure vessel) water-water energetic reactors (known as VVER in Russia) is regulated by the requirements of the Russian normative documents [1]. One of the key measures is the validation of calculated procedures by ex-vessel neutron activation measurements at VVER nuclear power plants. The neutron dosimetry research for every reactor component includes validation calculations by appropriate codes. The standard method was based on the three-dimensional synthesis method with input from results of calculations from the deterministic code DORT [2] coupled with the BUGLE-96T library [3]. It was identified in comparison between experiments and calculations that this method gives different results outside the active core. Hence, the usage of a three-dimensional code is essential. For this reason, the Scientific and Engineering Centre for Nuclear and Radiation Safety wants to use the Monte Carlo code TRAMO for their fluence calculations. Thus, the purpose of this work is the possible application of TRAMO for expert calculations of radiation load parameters of reactor components for all VVER reactor types, verification of results by comparison with other codes (deterministic and Monte Carlo), and validation by neutron activation measured data. The paper presents the main options of the new version of TRAMO for neutron fluence calculations suggested by the Scientific and Engineering Centre for Nuclear and Radiation Safety.