First Experience with the Consolidation of WWER Reactor Pressure Vessel Knowledge through a New Method

One of today¿s activities of the Joint Research Centre¿s (JRC) Institute for Energy (IE) concerns data management and dissemination in nuclear safety. An ¿Online Data & Information Network¿ (ODIN) is set-up, which maintains one document database and four engineering databases. These databases aim to deploy networks for energy related research & development, specifically for nuclear energy and to provide the public experimental data of European projects on mechanical and thermo-physical material properties in comparison with international standards and recommendations. Due to its long lasting experience and being in a key position as regards web based d-base (e.g. ODIN), the IAEA for example has recently transferred the reactor surveillance data-base to the IE. Lately, many stakeholders, such as Institutes, R&D Organisations, Regulators, Utilities, Governmental Organisations, have recognised the need for collecting, preserving, consolidating (validating), and disseminating nuclear knowledge (documents, competences and data), in order to make it easily accessible to future generations through modern informatics tools and training and education measures. A broad spectrum of components and technologies should be considered, i.e. reactor pressure vessel (RPV), piping, internals, steam generator, etc. regarding knowledge, material data and practices. In the long run, it will also support future decommissioning exercises of nuclear installations as a valuable knowledge source. In addition to the knowledge in each Member State, the IE produced a long standing record of results from its own institutional activities and even more through the participation to a large number of European Network partnership projects. It is important, besides preservation, to consolidate the enormous amount of scientific results produced since. Therefore, the IE has developed a method for consolidation of nuclear knowledge. The method relays on the mobilisation of all identified leading experts in the EU in re-evaluating old knowledge and consolidating what is necessary to create training materials for the new generations. This method was applied for a pilot study for consolidating and preserving WWER RPV safety related knowledge, which is scattered in many countries and in different languages, facing a serious issue in terms of getting lost. This initiative could be the start of a wider Nuclear Knowledge Preservation and Consolidation activity. Experience gained from the first exercise will be presented in this paper.JRC.F.4-Safety of future nuclear reactor

Annealing and Re-embrittlement of Reactor Pressure Vessel Materials - State of the Art Report

Annealing of a reactor pressure vessel embrittled by neutron irradiation constitutes the only known technique to restore the initial material properties, to an extent that depends on the annealing conditions and on the materials. This technique is used in WWER-440 type reactor pressure vessels. A very important related issue is the one of re-embrittlement behaviour of the material after the annealing. In this respect, there is an obvious link with radiation embrittlement understanding. This report compiles the vast amount of information on annealing and re-embrittlement, which is available in the European countries where such annealing operations have been performed. In addition this topic was also investigated in various TACIS-PHARE projects, and the conclusions are included here as well. To complete the state-of-the-art, the results from a number of annealing experiments carried out in US on Western type RPV steels have also been considered.JRC.F.4-Nuclear design safet

Project PISA: Phosphorus Influence on Steel Ageing

The integrity of the pressure vessel is vital to the safe operation of a nuclear reactor. It is therefore necessary to monitor or predict the changes in the reactor pressure vessel (RPV) material during operation. Exposure to irradiation (or elevated temperatures) causes the segregation of phosphorus to internal grain boundaries in RPV steels. This, in turn, encourages brittle intergranular failure of the material. The PISA project had the objective of reducing the uncertainties associated with the impact of this failure mechanism on the properties of the RPV, both during service and at the end-of-life. This report presents the experimental results on the segregation of P and C during irradiation and thermal treatments, and the associated mechanical property changes, generated within PISA. The new data cover a range of bulk P levels, irradiation temperatures and fluences, steel types and product forms. In all cases only modest increases of P level on the grain boundary have been observed in commercial steels. Segregation is higher in pre-strained than in unstrained material. In addition a model for P segregation under irradiation has been developed, and shown to be capable of fitting the experimentally observed changes in P level after irradiation. Significant insight into the development of the microstructure under irradiation has thereby been obtained. Overall, the data and modelling together indicated that relatively small amounts of segregation are likely to occur under most reactor operational conditions in homogeneous commercial steels, and an (unexpectedly) small amount of additional embrittlement likely to derive from this process during reactor service.JRC.F.4-Nuclear design safet

Assessment of a Collection of Papers on Property-Property Correlations in the Field of Radiation Embrittlement of WWER Reactor Pressure Vessel Steels

Knowledge management (NKM) tools continue to be enhanced to support information exchange and knowledge distribution. New tools for supporting NKM were initiated by international organizations, including a web-based knowledge portal, ODIN, developed by the JRC of the European Commission. This portal is dedicated to knowledge preservation and consolidation and includes a nuclear database where papers on WWERs are collected and electronically stored. A collection of around 100 papers selected from the ODIN database on property-property correlations for irradiated reactor pressure vessel steels has been reviewed in order to have a clear picture of the advances achieved so far in this field, and to determine existing gaps and open issues. It allows identifying research activities needed, as described in this paper. The review performed focus on two main areas. Namely, non-destructive examination of the embrittlement condition and fracture and Charpy toughness determination, including specimens size and constrain issues.JRC.DDG.F.4-Safety of future nuclear reactor

Certification Report of 15Kh2MFA/15Cr2MoVA Steel and its Welds for WWER Reactor Pressure Vessels

The main goal of this Certification is to summarize all available important information and experience on base and weld materials for WWER-440 reactor pressure vessels (RPV) on the basis for Knowledge Management in other countries and Preservation/Dissemination. The first of all the Report summarizes data which was received in Soviet Union (later Russia) and Czech Republic as the main countries which had produced this type of RPV. The book summarizes historical basis for the choice of materials, main principles of RPV design and manufacturing as well as results from Qualification Test Programmes, statistical evaluation of test results from individual type of semi-products/welds and final vessels. Information on defects/repair of semi-products/vessels was included together with available information about vessels behaviour during operation. Finally, comparison of existing data with Code/standards requirements will be given.JRC.DDG.F.4-Safety of future nuclear reactor

Study of Aging Mechanisms for Structural Materials within SAFELIFE Project

Abstract not availableJRC.F-Institute for Energy (Petten