Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the conventional positron annihilation lifetime set-up and the Pulsed Low Energy Positron System (PLEPS) was applied for the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels as well as for investigation of new materials foreseen for Generation-IV and fusion facilities. PAS results obtained in framework of several scientific projects are summarised and correlated to other techniques. Possibilities for experimental modelling of irradiation effects using ion implantation instead of neutron irradiation are discussed in the paper. The actually running program of micro structural study of Fe-Cr binary alloys is presented in detail. For the next effective study of radiation degradation of fusion as well as fission materials, the experimental micro structural study of model alloys in combination to computer simulations is recommended.
Activities connected to the nuclear knowledge preservation are ongoing in the EC-JRC Institute of Energy with the intention to collect all available information about reactor pressure vessels of WWER type reactors as well as to analyze and summarize the most important items and issues. This activity is in line of the European Community FP6 projects PERFECT (Prediction of irradiation damage effects on reactor components) and mainly COVERS (Coordinated action on WWER safety) in which all WWER operating countries also take part. Actually, the electronic database was created and is accessible for young or expired researchers in this area. The access is recommended via ODIN (Online Data and Information Network) https://odin.jrc.nl/doma. After registration you can enter the WWER DoMa-db: "Database of references for knowledge management and Preservation on WWER reactor pressure vessel". For the access to confidential information you have to ask an indicated administrator. The nuclear knowledge management is realized not only via database creation or education process during undergraduate (Bc.), graduate (MSc.) and postgraduate (PhD.) study but also via specialised training courses in a frame of continuous education system, research activities and projects, workshops seminars, ect. For illustration of the actual status and possibilities, the Slovak nuclear knowledge model is used. Unfortunately, decrease of number of employees in nuclear and "human ageing"of experts seems to be a serious problem not only on world but also in Slovakia.
Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system “Sipping in pool” delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand “SVYP-440” for monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system “Sipping in pool,” the limit values of leak tightness were established.
Nuclear grade fuel claddings act as a barrier against release of fuel particles into the coolant water during nuclear power plant operation, handling, and wet or dry storage of the spent fuel rods. The integrity of claddings is always a critical issue during the reactor operation, storage, or loss of coolant accidents (LOCA). After Fukushima disaster, cladding materials are widely studied, as the neutron transparent zirconium (Zr)-based alloys or accident tolerant fuel claddings, which should reduce the high-temperature oxidation rate of Zr-based alloys and enhance the accident tolerance. One part of the long-term investigation of Zr-based claddings in the Research Center Řež contributes to the characterization of material behavior after creep testing and high-temperature oxidation, i.e., conditions that may occur during reactor operation or storage of the spent fuel. Then, the microstructure response can be investigated by means of standard as well as advanced microscopy methods. In this study, we introduce the methodology of the cladding microstructure complex characterization demonstrated on the reference (nonirradiated) Zr-1%Nb alloy after creep testing. In the Zr-1%Nb microstructure, the hydrogen is present under some conditions as the hydride phases ZrH or ZrH2 that cause the mechanical properties degradation. As well, the hydride reorientation can pose a risk. Therefore, currently used methods of hydrides evaluation are presented. Furthermore, we introduce the microstructure analysis of the zircalloy-4 cladding after high-temperature oxidation at the temperature 1300 °C simulating LOCA. For this study, the new automatic chemical analysis procedure was developed. The main issue is the determination of the oxygen concentration and alloying elements behavior after critical increase of oxygen in the Zr-alloy tube at LOCA, also concerning the hydrogen content in the microstructure. Based on the element concentration gradients, the phase transitions from the oxide ZrO2 to the prior β-Zr can be determined and contributes to the phase diagrams experimental verification. In our contribution, the significant role of the electron microscopy methods “from micro to nano-scale” in the nuclear research is emphasized in these two research topics.
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