By now, in Russia more than 70 000 fuel assemblies (FA) have operated their service life at NPPs with RBMK–1000 (11 units); the discharged FAs have been stored in water of cooling ponds and in spent nuclear fuels (SNF) storage facilities for 25 years. During their operation, more than 3 000 spent fuel assemblies became leaky while claddings of some integer fuel rods were subjected to a substantial wear (fretting-corrosion up to 400μ in depth, nodular corrosion to a depth of 380μ — both kinds of the wear proceeded under spacer grids). In this case, the claddings revealed a higher hydrogen uptake and degradation of mechanical properties. During the wet storage in cans, spent FAs show general corrosion proceeding at the rate of 3–5 μ/ year. The paper points out that as applied to a long-term (up to 50 years) dry storage of RBMK-1000 spent nuclear fuel, consideration is to be given to the following degradation mechanisms proceeding in stainless steel fuel rod claddings and FA components: Oxidation of Zr-1%Nb alloy (fuel rod claddings, central channel) in the atmosphere of storage (air, nitrogen, argon and impurities). • Creep of a fuel rod cladding material. • Delayed hydride cracking (DHC) of a fuel rod cladding. • Corrosion induced cracking of FA components fabricated from stainless steels. The first results obtained in assessing the creep rate of standard fuel claddings effected by the internal pressure are under consideration.

This paper is devoted to some aspects of the physical and chemical evolutions of irradiated fuels in a closed system, which implies no mass transfers with the surrounding medium. This situation is not only relevant for dry interim storage, but also for the first stages of a geological disposal scenario, prior to the re-saturation of the near field and the access of water to the spent fuel. Some potential evolutionary mechanisms are presented and discussed: • The chemical evolution of spent fuel: natural radioactive decay leads to the formation of elements with different oxidation states and consequently, could modify the oxygen potential of the irradiated fuel. Thermo-chemistry calculations performed for decay periods of up to 300 years, show however that the change in composition of spent UOX fuel is not sufficient to significantly affect the oxidation state of the oxide matrix. The temperature dependency of oxygen potential is not modified with time, and therefore the oxygen potential will decrease with the temperature decrease in the long term. • Helium behaviour in the fuel has been identified as a major issue since little is known on the mobility of helium in uranium oxides or mixed oxides or on the potential consequences of helium release on long term fuel rod integrity. A comparative assessment of the amount of helium generated in standard UOX and MOX fuels, at the same rod average burnup is presented for storage periods of up to 10,000 years. More helium is generated in MOX fuels than in UOX fuels due to their higher content in alpha emitters. Potential consequences for fuel storage are discussed based on two extreme hypotheses whereby helium is entirely released or retained within the oxide. A preliminary analysis of existing data on helium solubility, its diffusion in uranium dioxides and reactor feedback is presented and reveals that uncertainties make it difficult to predict with confidence the extent to which helium release could occur under storage conditions or if indeed the integrity of the pellet could be undermined in the long term.

The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd 2 O 3 ) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods. Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated. Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3 ® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3 ® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3 ® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3 ® cask is designed for transport and the intermediate storage of at least 50 years. Although a defect of the leaktightness of a CASTOR BR3 ® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3 ® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.

In this paper, some results of microstructure analyses on spent MOX fuel with burnup of about 20 GWd/tM, stored for an extended period, are presented. The samples are taken from fuel that has been stored for 25 years either in intact cladding or exposed to dry air. The results of this investigation are compared to the microstructure observations made in the post-irradiation campaign, conducted on the same or sibling fuel rods shortly after reactor unloading (i.e. about twenty years ago). The fuel from the long term stored intact rods demonstrated an identical behaviour, compared to its state shortly after reactor unloading. Segments of fuel rods that were exposed to dry air showed an alteration of the grain boundary cohesion of the UO 2 grains, while apparently the Pu-rich zones were unaffected. These findings are coherent with the results found in accelerated ageing experiments. On the other hand, the formation of secondary phases with higher oxidation states of uranium, often observed during accelerated ageing experiments, could not be confirmed here.

The balance between the amount of plutonium produced by reprocessing and recycled in the PWR leads EDF, the French Utility, in cooperation with CEA to investigate the feasibility of dry interim storage for the spent fuel assemblies, especially the MOX ones. The safety criteria require to prevent the cladding from any loss of integrity. The thermal creep of the cladding due to the internal fuel rod pressure is assumed to be the most relevant mechanism susceptible to lead to failure. Therefore, the simulation of the creep behavior for quite long term dry storage and under variable loading conditions (temperature and internal rod pressure) requires a creep law integrating the suitable parameters. Using a step-by-step approach based on available data base on as-received and irradiated CWSR Zircaloy 4 cladding, a creep law under biaxial loading has been developed, this paper presents the creep state equation.

Uranium fuel rods were produced in the nuclear fuel site. The buildings should be dismantled after decontamination and the site should be released for industrial use. The individual dose to the critical group is limited to an annual value of about 10 μSv. The determined specific activity for remediation of the site was a mean value of 60 mBq/g total activity. For the building rubble and soil primarily two pathways, disposal at a landfill and refill of a disused salt mine, were considered. As a result of the investigations the total activity for the disposal at a landfill had to be limited to about 6,6 GBq. For the refill of the salt mine the estimated individual dose fell below the dose limit in the range of 10 μSv/y.

The United States Department of Energy (DOE) Office of Civilian Radioactive Waste Management conducted spent nuclear fuel integrity and cask performance tests from 1984–1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). Between 1994 and 1998, DOE also initiated a Spent Fuel Behavior Project that involved enhanced surveillance, monitoring, and gas-sampling activities for intact fuel in a GNS CASTOR V/21 cask and for consolidated fuel in a Sierra Nuclear VSC-17 cask. The results of these series of tests are reported in this paper. Presently, DOE is involved in a cooperative project to perform destructive evaluations of fuel rods that have been stored in the CASTOR V/21 cask. The results of those evaluations are presented elsewhere in these proceedings in a paper entitled “Examination of Spent PWR Fuel Rods after 15 years in Dry Storage”.

Potential damage mechanisms postulated for spent fuel in dry storage include stress corrosion cracking (SCC), delayed hydride cracking (DHC), and accelerated (tertiary) creep leading to creep rupture. The primary conditions that govern the evolution of these damage mechanisms during dry storage are the spent fuel rod’s internal chemical environment and its thermal and mechanical histories. In a recent publication, (see cited references), thorough evaluation of these mechanisms, utilizing conservative estimates of pre-existing thermomechanical and physical conditions of the cladding, show that SCC and DHC are not likely operative mechanisms in dry storage. This leaves creep rupture as the primary mechanism of concern. Evaluation of this mechanism is the subject of the present paper. Results of creep modeling and analysis are presented to show that self-limiting creep deformations govern the behavior of fuel rods in dry storage. This leads us to suggest that a liberal strain limit of several percent, with a side condition on stress evolution to remain below the local yield strength, is an appropriate acceptance criterion for dry storage.

The oxidation of nuclear fuel has been studied for over 40 years and some engineering criteria has been established to evaluate the evolution of a defective UOX fuel rod in contact with air in order to assist the designer of an (interim) storage facility. Within the French context, MOX fuel also has to be taken into consideration. In the PRECCI program, some oxidation tests were conducted on UOX and MOX samples, non-irradiated and irradiated, to compare and to analyse the difference between these two types of fuel. The weight gain curves show a different behaviour depending on irradiation: non-irradiated MOX is less oxidised than UOX, but irradiated MOX oxidises faster than UOX. This difference is explained by the fast oxidation of irradiated UO 2 grain boundaries, because of which irradiated UO 2 ceramics behave as a powder from an oxidation point of view. The weight gain curves for irradiated samples were treated to determine the oxidation rate constant k, which was found consistent with the values reported in the literature. Yet our results tend to prove that k depends on burn-up. The existence of an oxidised layer in the UO 2 grains was confirmed with the examination of samples taken during the oxidation test. Optical observation revealed a new phase on the grain edge which was proved to be oxygen enriched by EPMA. Although the oxidation of plutonium enriched clusters was proved with EPMA, it was not possible to characterise it with optical observations. In order to better understand the relationship between oxidation kinetics determined by weight gain measurements and actual fuel rod degradation, an attempt was made to estimate the gas fluxes in a defective fuel rod. The gas propagation inside the fuel rod was estimated at room temperature by interpreting fuel rod piercing data. The gas flux inside a fuel rod was then compared to the oxygen consumption by oxidation and to the gas supply through the defect. It was concluded that oxygen is readily consumed by oxidation and does not have the ability to propagate in the fuel rod. A stop and go degradation process for the evolution of a defective fuel rod is thus proposed based on this assumption.

Utilities worldwide are using dry-cask storage systems to handle the ever-increasing number of discharged fuel assemblies from nuclear power plants. In the United States and possibly elsewhere, this trend will continue until an acceptable disposal path is established. The recent Fukushima nuclear power plant accident, specifically the events with the storage pools, may accelerate the drive to relocate more of the used fuel assemblies from pools into dry casks. Many of the newer cask systems incorporate dual-purpose (storage and transport) or multiple-purpose (storage, transport, and disposal) canister technologies. With the prospect looming for very long term storage — possibly over multiple decades — and deferred transport, condition- and performance-based aging management of cask structures and components is now a necessity that requires immediate attention. From the standpoint of consequences, one of the greatest concerns is the rupture of a substantial number of fuel rods that would affect fuel retrievability. Used fuel cladding may become susceptible to rupture due to radial-hydride-induced embrittlement caused by water-side corrosion during the reactor operation and subsequent drying/transfer process, through early stage of storage in a dry cask, especially for high burnup fuels. Radio frequency identification (RFID) is an automated data capture and remote-sensing technology ideally suited for monitoring sensitive assets on a long-term, continuous basis. One such system, called ARG-US, has been developed by Argonne National Laboratory for the U.S. Department of Energy’s Packaging Certification Program for tracking and monitoring drums containing sensitive nuclear and radioactive materials. The ARG-US RFID system is versatile and can be readily adapted for dry-cask monitoring applications. The current built-in sensor suite consists of seal, temperature, humidity, shock, and radiation sensors. With the universal asynchronous receiver/transmitter interface in the tag, other sensors can be easily added as needed. The system can promptly generate alarms when any of the sensor thresholds are violated. For performance and compliance records, the ARGUS RFID tags incorporate nonvolatile memories for storing sensory data and history events. Over the very long term, to affirmatively monitor the condition of the cask interior (particularly the integrity of cover gas and fuel-rod cladding), development of enabling technologies for such monitoring would be required. These new technologies may include radiation-hardened sensors, in-canister energy harvesting, and wireless means of transmitting the sensor data out of the canister/cask.

The “direct disposal of spent fuel” as a part of the current German reference concept was developed as an alternative to spent fuel reprocessing and vitrified HLW disposal. The technical facilities necessary for the implementation of this part of the reference concept, the so called POLLUX ® concept, i.e. interim storage buildings for casks containing spent fuel, a pilot conditioning facility, and a special cask “POLLUX” for final disposal have been built. With view to a geological salt formation all handling procedures for the direct disposal of spent fuel were tested aboveground in full-scale test facilities. To optimise the reference concept, all operational steps have been reviewed for possible improvements. The two additional concepts for the direct disposal of SF are the BSK 3 concept and the DIREGT concept. Both concepts rely on borehole emplacement technology, vertical boreholes for the BSK 3 concept und horizontal boreholes for the DIREGT concept. Supported by the EU and the German Federal Ministry of Economics and Technology (BMWi), DBE TECHNOLOGY built an aboveground full-scale test facility to simulate all relevant handling procedures for the BSK 3 disposal concept. GNS ( Company for Nuclear Service ), representing the German utilities, provided the main components and its know-how concerning cask design and manufacturing. The test program was concluded recently after more than 1.000 emplacement operations had been performed successfully. The BSK 3 emplacement system in total comprises an emplacement device, a borehole lock, a transport cart, a transfer cask which will shuttle between the aboveground conditioning facility and the underground repository, and the BSK 3 canister itself, designed to contain the fuel rods of three PWR-fuel assemblies with a total of about 1.6 tHM. The BSK 3 concept simplifies the operation of the repository because the handling procedures and techniques can also be applied for the disposal of reprocessing residues. In addition to this, the feasibility of the direct disposal of transport and storage casks, the so-called “DIREGT concept”, is being investigated. The implementation of this concept would avoid the necessity to separate fuel rods from structural parts and to procure custom-made final disposal casks. All investigations and studies performed so far support the feasibility of direct disposal of spent fuel in multipurpose casks as today used for transport and storage. Both additional concepts, BSK 3 and DIREGT, are expected to simplify disposal processes and to reduce operational risk without any compromise in long-term radiological safety aspects.

UO 2 containing different fractions of short-lived alpha-emitters, the so-called alpha-doped UO 2 simulates the level of activity of spent fuel after different storage times, and can be used to study the effects of radiolysis on the corrosion behaviour of aged spent fuel exposed to groundwater in a geologic repository. Furthermore, the integral over time of the alpha-decay in alpha-doped UO 2 can simulate the decay damage accumulated in spent fuel during storage. This allows investigating property modifications occurring to the fuel during storage periods of interest (e.g. in view of spent fuel retrieval or in view of final disposal) within a laboratory-acceptable timescale. Periodical measurements of lattice parameter are performed on high activity alpha-doped UO 2 to investigate the build-up of radiation damage and evaluate possible dose rate effects. Additionally, annealing tests combined with He-release measurements using a Knudsen cell and with microstructure examination using TEM are performed to establish a correlation among the annealing of damage in the microstructure (mainly characterized by dislocation loops) and the release behaviour of He. The effects on the microstructure due to the accumulation of He and α-decay damage are of interest as they may considerably affect the mechanical integrity of the fuel rods, by causing e.g. swelling or cracking in the fuel and/or overpressurization of the cladding. Alpha-doped UO 2 with specific activities spanning over three orders of magnitude and undoped UO 2 were used in static leaching experiments at room temperature in deionized water under nominally anoxic conditions. Under these experimental conditions (single effect studies) a clear dissolution enhancing effect of alpha-radiolysis was observed coupled with the establishment of higher redox potential due to the radiolytic process. An alpha-activity dependence of the dissolution behaviour was observed.