Nuclear Fuel Cycle and Materials Section
Nuclear Power Reactor Fuel Engineering
Supporting the sharing of experience in the development and use of fuel structural materials and water chemistry management in nuclear power plants
Objective: To maintain and improve in-pile characteristics of fuel, core and primary circuit components for ensuring nuclear fuel cycle; to increase understanding of the mechanisms of irradiation-assisted degradation of properties and characteristics of the above-mentioned materials with taking into account environmental (coolant chemistry, hydrogen, etc) impact.
Responsible Officers: Victor INOZEMTSEV and Pablo ADELFANG
Details: All Member States with nuclear power programmes have an interest in the degradation of reactor core and primary circuit materials. The degradation of the mechanical and physical properties of these materials, often enhanced by irradiation and the pickup of hydrogen or other species from the coolant, controls the effective lifetimes of fuel assembly, control rod and primary circuit materials. Consequently, lifetime extensions of power reactors rely on increasing knowledge of the mechanisms controlling the behaviour of ageing of irradiated materials. To coop with this, the above-mentioned phenomena are to be studied and understood, and respective remedies (e.g. radiation and corrosion-resistant clads, core and primary circuit materials, appropriate water chemistry regimes, etc.) are to be developed and proposed to utilities.
Within this project, there are two running CRPs, namely on “Data processing technologies and diagnostics of water chemistry and corrosion control in nuclear power plants (DAWAC, 2001-2005) and on “Hydrogen and hydride induced degradation of the mechanical and physical properties of zirconium-based alloys” (1998-2003).
A progress in the DAWAC CRP was evaluated at the 2nd meeting held in November 2002. The project, joined by 17 organizations from 16 countries, is directed on development of recommendations to NPP operators on optimal combination of coolant on-line monitoring techniques and off-line laboratory measurements to keep corrosion and dose rates on minimally achievable level. It was confirmed that experiences gained in laboratory tests and campaigns at plants should be incorporated to existing data collecting and diagnostic systems. Participants came to the conclusion that emphasis should be given to smaller scale Semi-Expert Systems, which are rather cheaper and simpler to operate than full scale Expert Systems. In the area of water chemistry monitoring and control, the amount instrumentation needed in the future was discussed. The general consensus was that not too much because that would lead to too many data points to be followed and then again not too few, because one does not want to compromise the accuracy of the water chemistry control. First observations have been summarized for different types of reactors (BWRs, PHWRs, PWRs, RBMKs and WWERs). With regard to the need for development of some new monitoring techniques, like Li monitor being developed at EdF, sensor to be used directly in steam generator crevice, etc, more detailed discussion will be carried out during the next meeting.
A study of the delayed hydride cracking (DHC) of the zirconium alloys used in pressurized heavy water reactors (PHWRs) involving 10 countries was completed within the framework of a Co-ordinated Research Project (CRP). Many of the results have been published already in scientific papers and the CRP Report is in preparation for publication. This CRP achieved very effective transfer of know-how at the laboratory level in three technologically important areas: 1) controlled hydriding of samples to predetermined levels; 2) accurate measurement of hydrogen concentrations at the relatively low levels found in pressure tubes and RBMK channel tubes; and 3) in the determination of DHC rates under various conditions of temperature and stress. DHC has led to the failure of pressure tubes in CANDU reactors and is of serious concern to the operators of all PHWRs and may also contribute to fuel cladding failure in all types of water-cooled reactors. An experimental technique to study short sections of fuel cladding tubes was developed by one of the participating laboratories and it is anticipated that detailed research programme using the technique to elucidate axial slitting of fuel cladding will form the subject of a new follow-up CRP. Final report is under preparation.
Upon recommendation of the TWGFPT, new activity will start in 2003 with a purpose to analyze impact of technological factors and irradiation on mechanical properties of Zr-based alloys. It is planned to finalize a TECDOC on the subject in 2003.
In 2002 assistance was provide to the IAEA NS Department in preparation of a TECDOC on “Comparison of fuel safety criteria between WWER and Western PWRs”. Document will be published in 2003.