I.L.F. Ray

Transmission electron microscopy observation on irradiation-induced microstructural evolution in high burn-up UO2 disk fuel

In order to identify the conditions of the rim structure formation as a function of burn-up and temperature, and to clarify the formation mechanism of this restructuring, UO2 fuel disks were irradiated at four thermal conditions, between 400 and 1300 °C, and at four different burn-ups, between 36 and 96 MWd/kgU, without external mechanical constraint. The microstructural evolutions as a function of the irradiation parameters are observed by high resolution scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The SEM observations reveal the transition from original to sub-divided grains of rim structure and make clear that the burn-up threshold is between 55 and 82 MWd/kgU. The temperature threshold of this restructuring could be 1100±100 °C. Moreover, polyhedral sub-divided-grains with size ranging between 0.5 and 2 μm, not only rounded grains in the size range 150–350 nm, are clearly observed. These configurations are explained by assuming that the grain sub-divisions occurred homogeneously within the original polyhedral grains, while the existence of rounded grains might be due to free surface effects. TEM observations of re-structured samples show that most of sub-grain boundaries are low angle and are heavily decorated by fission gas bubbles in the range 3.5–8 nm. In the non-restructured samples, dislocations and small precipitates are present, and many of the bubbles form “strings” along dislocation lines. In specimens irradiated at high temperature, many dislocations seem to be anchored by fission product precipitates. These results suggest that the formation mechanism of the restructuring is based on polygonization, and the precipitates could have some “pinning effect” on dislocations and defect clusters.

Transmission electron microscopy study of fission product behaviour in high burnup UO2

Abstract Transmission electron microscopy (TEM) has been used in an extensive study of the microstructure of unirradiated, high burnup irradiated and transient-tested samples of LWR nuclear fuels. The object of the study was to determine the behaviour of the solid and gaseous fission products under a variety of conditions, and their interactions with the fuel microstructure. Quantitative determinations were made of fission gas bubble densities and size distributions, the related local microswelling and the average dislocation densities at a number of radial positions. Some of the implications of these results on fuel behaviour under steady state and transient conditions up to about 4.5% FIMA burnup are mentioned. Some observations relating to the change in fuel microstructure at the pellet rim are discussed.

Oxide fuel transients

Abstract The behavior of UO 2 fuel from power reactors has been studied up to high burn-up (~ 60 GWd/tU), under both steady state and power transient conditions in the range of 35 to 48 kW/m. Careful post-irradiation examination involving electron probe microanalysis, transmission and scanning electron microscopy in addition to detailed hot cell examination have provided a large data base on the radial migration and release of volatile fission products. Theoretical evaluation with different computer codes and supporting laboratory experiments established the basis for understanding the kinetics and mechanisms operative during transients in high burn-up UO 2 . Typical examples are given to demonstrate the degree of understanding achieved

An electron microscopy study of the RIM structure of a UO2 fuel with a high burnup of 7.9% FIMA

Transmission and high resolution scanning electron microscopy were used to analyze the microstructure of the periphery of a UO2 pellet irradiated to a cross-section average burnup of 7.9% FIMA, with a period of increased temperature at about half the final burnup. Local burnups at the pellet surface reached nearly 23% FIMA. In this rim region the original grains of about 10 μm diameter were subdivided into about 104 subgrains of 0.15 to 0.30 μm diameter. The spread in subgrain orientation was small ( 1 mm. The size of the subgrains was found to be largely independent of depth.

Local fission gas release and swelling in water reactor fuel during slow power transients

Abstract Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r / r 0 ~ 0.75. Release tunnels were fully developed at r / r 0 ~ 0.55 with the result that gas release was 60–70% at this position.

Dense fuels in Europe

Abstract After a reduction of the activities on dense fuels in Europe during the late seventies, the interest in this fuel type was revived at the beginning of the eighties, when the economy of the closed FBR fuel cycle was reassessed. Strategies were developed to bring its economy into equilibrium with that of the LWR fuel cycle. The most important step was the decision to develop the European Fast Reactor (EFR) and to continue using oxide fuel at first. However, for economical reasons in an optimised fuel cycle a better fuel than the oxide should be employed. This fuel must avoid the deficiencies of the oxide but retain its advantages. Mixed nitride would satisfy this condition if it can be demonstrated that it can attain a burnup of at least 15 at%. With the know-how available in Europe it should be possible to achieve this goal with a rather limited research programme during the coming 10 years.

The Sodium-Bonding Pin Concept for Advanced Fuels Part I: Swelling of Carbide Fuel up to 12% Burnup

Analyse detaillee du comportement de combustible hyperstœchiometrique (U,Pu)C a liaison sodium pour des taux de combustion de 2,8 a 12,5% dans le reacteur Rapsodie. Determination des quatre contributions au gonflement du combustible en fonction de la temperature et du taux de combustion

Fission Gas Behaviour during Power Transients in High Burn-Up LWR Nuclear Fuels Studied by Electron Microscopy

A Transmission Electron Microscope study has been made of UO2 nuclear fuel samples which have been subjected to short term in-reactor power transients, involving an increase of the fuel temperature of about 300°C. Under steady state operating conditions most of the fission gas (Xe, Kr) is retained in solution in the oxide fuel matrix, or is precipitated into a population of very small fission gas bubbles (<3nm diameter). This bubble population is continuously subjected to re-solution and re-nucleation, and large fission gas bubbles are not able to grow. In contrast, in fuels with about 4.5% FIMA burn up, containing about 0.5 at% Xe and 0.06 at% Kr as fission gases, subjected to an increase in temperature through a power transient, a population of large fission gas bubbles grows, which is almost invariably associated with the dislocation networks. The implications of the growth of this fission gas bubble population is discussed.

Internal conversion in energy dispersive X-ray analysis of actinide-containing materials.

The use of X-ray elemental analysis tools like energy dispersive X-ray (EDS) is described in the context of the investigation of nuclear materials. These materials contain radioactive elements, particularly alpha-decaying actinides that affect the quantitative EDS measurement by producing interferences in the X-ray spectra. These interferences originating from X-ray emission are the result of internal conversion by the daughter atoms from the alpha-decaying actinides. The strong interferences affect primarily the L X-ray lines from the actinides (in the typical energy range used for EDS analysis) and would require the use of the M lines. However, it is typically at the energy of the actinides M lines that the interferences are dominant. The artifacts produced in the X-ray analysis are described and illustrated by some typical examples of analysis of actinide-bearing material.

Tracing the Origin of Diverted or Stolen Nuclear Material through Nuclear Forensic Investigations

In the last fifteen years we have see the emergence of a new and potentially hazardous form of smuggling: that of nuclear and radioactive materials. This triggered the development of a new discipline in science, enabling to support law enforcement authorities in combating illicit trafficking and dealing with criminal environmental issues: nuclear forensics. Existing analytical techniques as used in material science, in nuclear material safeguards and in environmental analysis, were adapted to the specific needs of nuclear forensic investigations. Characteristic parameters (e.g. isotopic composition, chemical impurities, macro- and microstructure) can be combined to a “nuclear fingerprint”, pointing at the origin of the material. Further research is being carried out, aiming at identifying other useful material characteristics in order to reduce the ambiguities often remaining in the interpretation of the data and in the source attribution. New methodologies need to be developed, validated and implemented in order to determine parameters with good precision and accuracy. The availability of up-to-date reference on nuclear material is essential in order to identify the origin and the intended use of the material, or to exclude certain origins.