Volume 83, Issue 6

The research on safety of dry stored nuclear fuel poses a number of challenges related to understand and predict its performance under the anticipated conditions. The ultimate goal is the development of analytical capabilities for predicting the integrity of the fuel rod. In this regard, the research carried out by the Unit of Nuclear Safety Research of CIEMAT is enabling the use of predictive tools and analysis methodologies whereby safety conditions can be assessed. To do so, the experience gained in fuel rod thermal-mechanics in reactor is used as a basis; particularly, the fuel performance code FRAPCON is being extended to the dry storage technology, called FRAPCON-xt. This paper shows the work carried out so far, including some applications on aspects related to cladding embrittlement due to hydrides as one of the main concerns.

As a common part of nuclear waste management strategies, used nuclear fuel is dry stored in casks. After a sufficient time of wet cooling in fuel pools, the fuel assemblies in Germany are dry stored in dual purpose casks mostly on-site until a deep geological repository will be available in the future. In the light of the extended and preliminary not foreseen timescales of up to 100 years of dry storage, open questions arise concerning the long term fuel rod behaviour. Most of these questions are linked to safety criteria e.g. the integrity and fuel encapsulation. Due to the variety of used nuclear fuels and cladding materials in Germany and the additional use of MOX fuels and very high burn-up values of 65 GWd/tHM averaged over a fuel assembly, some of the boundary conditions are unique and specific to the German situation. The GRS has a strong research program to investigate the long term behaviour of fuel assemblies under the German specific conditions of dry storage. Some aspects are presented in the following article.

The paper at hands discusses the development of an integrated method to predict the fuel rod behavior of LWR-fuel during extended dry storage scenarios. It is well known that degradation effects exist under certain circumstances and that a fuel rod integrity assessment for long-term storage, handling and transport must take these effects into account. It will be elaborated why a complete coverage of the whole lifetime of the fuel rod is necessary. This includes the fabrication, reactor operation, wet storage, drying process, and the dry storage itself. Regarding this approach, a discussion on open questions and challenges will be given.

Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. In this paper, the mechanical behavior of high burn-up spent fuel assemblies (> approx. 50 GWd/t HM , value averaged over the fuel assembly) under transport conditions is analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. Additionally, the gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated. Considerations and knowledge gaps for the transport after extended interim storage are issues of growing interest. In this context, practical approaches are discussed based on the experience of BAM within the safety assessment of packages approved for transport of spent nuclear fuel.

Absorbed hydrogen degrades the mechanical properties of zirconium alloys used for nuclear fuel claddings. Not only the total amount of hydrogen absorbed in the cladding tube but also the zirconium hydride orientation and its distribution influence the toughness of the material. For instance, the so-called delayed hydride cracking is caused by the diffusive re-distribution of hydrogen into the dilative elastic strain field ahead of crack tips. The paper presents in-situ and ex-situ neutron imaging investigations of hydrogen uptake, diffusion and distribution in zirconium alloys used for claddings. An overview about results of in-situ experiments studying the hydrogen uptake in strained Zircaloy-4, as well as ex-situ investigations of the diffusion of hydrogen in cold rolled Zircaloy-2 and Zr-2.5 % Nb alloy depending on temperature, rolling direction and thermal treatment and of the hydrogen re-distribution in the β-phase of Zircaloy-4 during a Three-Point-Bending-Test at 600 °C are presented.

During some operations previous to the dry storage, which involve high temperature and hoop stress on the cladding, a part of the hydride platelets may precipitate along the radial direction. As a consequence, cladding embrittlement may occur. In this paper, the experimental techniques developed for the authors to produce hydride reorientation in unirradiated cladding are shown. Samples of fuel cladding with 150 and 500 ppm of hydrogen were prepared. Subsequently, these samples were subjected to reorientation treatments by applying high temperature (400 °C) and constant hoop stress (60, 90, 120 and 140 MPa). Finally, ring compression tests (RCT) were carried out to study the cladding embrittlement. Samples after the reorientation treatment (i.e. with radial hydrides) were subjected to RCT at 20, 135 and 300 °C and the results are discussed.

This paper presents the methodology applied for the experimental and numerical investigation of the mechanical response of spent nuclear fuel rods under static and dynamic loads. The experimental activities were conducted at the JRC Karlsruhe where a 3-point bending test device and an impact tower have been developed and commissioned at the hot-cell facilities. Results are provided for two PWR samples. Load-displacement curves describe the mechanical response of the sample in the 3point bending tests, whereas an image analysis methodology has been developed to comprehend the sample's behaviour under dynamic loads (recorded using a high-speed camera). Finite Element Analysis (FEA) are used to simulate the rod's response based on static and transient structural models in ANSYS Mechanical.

Extended dry storage of spent nuclear fuel is a relevant issue in many countries operating nuclear power plants. Beside regulatory and security aspects there are questions with respect to the long-term integrity of the spent fuel as this is of relevance for final transportation and reloading to final waste repository casks. Within the frame of the BMWi project DCS-MONITOR, we investigate the potentials of different methods for non-intrusive monitoring of dry cask storage containers with spent nuclear fuel. These are thermography, radiation-based methods, and acoustic methods. For all of them we study the sensitivity and cross-sensitivity with respect to defined changes in the nuclear fuel distribution inside the containers. The analyses are mainly based on numerical simulations but also include some dedicated experimental studies.