Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been $\text{concluded}$ that the ratio of condensed, chemically-associated Cs and Te in the interface region,?s:Te, which in turn determines the Te activity, is $\text{controlled}$ by an equilibrium reaction between Cs₂Te and the oxide fuel, and that the value of ?s:Te is, depending on fuel 0:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (4̃:1) and FPLME (2̃:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.     

Ultrasonic inspections of fuel alignment pins

As a remedy to the practical problem of defects in fuel alignment pins made of Inconel X750, an inspection technique has been developed which fully meets the requirements of detecting defects. The newly used fuel alignment pins made of austenite are easy to test and therefore satisfy the necessity of further inspections.For the fuel alignment pins of the upper core structure a safe and fast inspection technique was made available. The inspection sensitivity is high and it is possible to give quantitative directions concerning defect orientation and depth. After the required inspections had been concluded in 1989, a total of 18 inspections were carried out in various national and international nuclear power plants in the following years. During this time more than 6000 fuel alignment pines were examined.For the fuel alignment pins the inspection technique provided could increase the understanding of the defect process. This technique contributed to the development of an adaptive and economical repair strategy.     

Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 10²⁶ n/m² as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.     

Diametral strain of fast reactor MOX fuel pins with austenitic stainless steel cladding irradiated to high burnup

The C3M irradiation test, which was conducted in the experimental fast reactor, “Joyo”, demonstrated that mixed oxide (MOX) fuel pins with austenitic steel cladding could attain a peak pellet burnup of about 130 GWd/t safely. The test fuel assembly consisted of 61 fuel pins, whose design specifications were similar to those of driver fuel pins of a prototype fast breeder reactor, “Monju”. The irradiated fuel pins exhibited diametral strain due to cladding void swelling and irradiation creep. The cladding irradiation creep strain were due to the pellet-cladding mechanical interaction (PCMI) as well as the internal gas pressure. From the fuel pin ceramographs and ¹³⁷Cs gamma scanning, it was found that the PCMI was associated with the pellet swelling which was enhanced by the rim structure formation or by cesium uranate formation. The PCMI due to cesium uranate, which occurred near the top of the MOX fuel column, significantly affected cladding hoop stress and thermal creep, and the latter effect tended to increase the cumulative damage fraction (CDF) of the cladding though the CDF indicated that the cladding still had some margin to failure due to the creep damage.     

Interactions between fuel pins and assembly components

The structure of a fuel element has the primary function of holding fuel pins in a regular array throughout their irradiation life, and secondary functions of permitting loading and unloading operations and transportation to be carried out without damage. Differential thermal expansions, irradiation-induced dimensional changes and flow-induced phenomena have to be accommodated. These factors can lead to fuel pin damage and ultimate loss of cladding integrity, or to distortions which may affect the thermal performance either in normal operation or in a loss-of-coolant accident. This paper discusses the various types of interaction that have been experienced and their consequences, as well as the design principles that should be followed to avoid them.     

On the failure of fast reactor fuel pins

The prediction of the timing and position of fuel pin failures is an important task in the modelling of fast reactor fuel behaviour. The range of processes that can provoke failure of fast reactor fuel pins in normal operating conditions and during hypothetical accidents is reviewed. Some of the mechanisms of failure are examined in more detail and the effect of hot spots and local stress concentrations is discussed. A review of failure criteria used in fast reactor fuel pin codes is given elsewhere, but the difficulties in applying various types of criteria are examined. Some discussion is also given on probabilistic approaches. Recommendations are given for a future approach to the problem of failure prediction, resolving the dilemma between inadequate empirical criteria and over-complex physically based approaches.     

Attack upon the cladding of oxide fuel pins by fuel and fission products

In parallel with post-irradiation examinations, a comprehensive out-of-pile experimental programme has been performed to determine the most important fission product reactions with four austenitic stainless steels at different oxygen potentials. Single as well as groups of fission products (simulated burn-up systems) have been used. Only the elements cesium, iodine and tellurium cause dangerous reactions with the cladding of an oxide fuel pin. The others are either not reactive or produced in such small quantities that their attack on the cladding is insignificant. Molybdenum is often found in the reaction zone of an irradiated oxide pin. However, according to our out-of-pile results it does not look as if molybdenum is a dangerous fission product. A decisive factor for the occurrence of reactions with the cladding is the oxygen potential in the fuel pin. As long as the O/M ratio of the fuel is markedly below 2.00, there are no dangerous reactions, neither with cesium nor with tellurium and iodine. The post-irradiation investigations (burn-up 1 to 10 at %) have shown that the cladding attack below 750 °C is most dependent on the inner wall temperature. Other factors, including fuel density, rod power and burn-up, seem to play a minor role. A noticeable reduction of the cladding attack was observed when the initial O/M ratio of the fuel was less than 1.98. A kinetic evaluation of some of the reactions observed in the out-of-pile tests has been attempted. At temperatures above 700 °C, the influence of temperature decreases markedly and the fission product concentration in the fuel becomes more important. There are indications that this also holds true for in-pile conditions.     

The development of grain-face porosity in irradiated oxide fuel

The predominant mode of fission gas release occurs through atomic diffusion to the grain boundaries. In oxide fuels the fission gases initially precipitate as an array of small lenticular bubbles of circular projection. The arrival of additional gas and vacancies causes these bubbles to grow and coalesce into fewer, larger bubbles. Depending on the irradiation conditions and temperatures, these bubbles may develop either as circular lenticular pores or as extended multi-lobed pores. Eventually the pores may intersect the grain edges where pathways may be formed which enable the gas to migrate to the outer geometry of the fuel and hence to the gap and the pin free volume. Recent extensive PIE campaigns on irradiated fuels have provided a large database of inter-granular porosity development and, from these, models of bubble growth, coalescence, morphological relaxation and venting have been developed.     

Plutonium redistribution in fast reactor mixed oxide fuel pins

Some redistribution effects of uranium and plutonium, caused by thermal diffusion and evaporation-condensation processes in mixed oxide fuels, are discussed by means of autoradiographs of sections of fuel pins irradiated in the fast flux of the RAPSODIE reactor. The change in the stoichiometric state as a function of burnup and the radial distribution of oxygen are described and their influence on the redistribution processes is discussed. A model and suitable data are given to calculate redistribution effects on the basis of thermal diffusion in fast reactor fuels. In fuel pins with power ratings of 500 W/cm and 600 W/cm the enrichment of plutonium around the central cavity produces an increase in the central temperature of about 100°C and 250° C, respectively.     

Cesium-uranium-oxygen chemistry in uranium-plutonium oxide fast reactor fuel pins

The Cs-U-O system has been reinvestigated in light of recently reported thermodynamic data for Cs₂U₄O₁₂ and recent phase data showing the existence of a new Cs-U-O compound with an $\text{O}\text{(U + Cs)}$ atom ratio less than that of Cs₂UO₄. Our experiments have confirmed the existence of a new phase, allowed the formula Cs₂UO_3.56 to be assigned, and generated thermodynamic data for this new compound. This new phase exists only at oxygen potentials that are too negative to be encountered in uranium-plutonium oxide fast reactor fuel pins. The compound Cs₂UO₄ appears to be the most likely one to be formed, with the formation occurring at the fuel-blanket interface.     

Thermophysical properties of irradiated uranium-zirconium fuel

The dependence of the thermophysical properties of metallic nuclear fuel — the alloy Zr-40U — in a wide temperature range on the amount of fission products accumulated is presented. Non-irradiated and irradiated samples with different degree of accumulation of fission products — 0.4, 0.6, and 0.9 g/cm³ — are investigated. The specific heat is measured in the range 50–1000°C, the temperature diffusivity is measured in the range 300–1000°C, and the variation of the dimensions and density of the samples on heating is also investigated. The thermal conductivity in the range 50–1000°C is calculated on the basis of the experimental data. __________ Translated from Atomnaya énergiya, Vol. 108, No. 1, pp. 6–9, January, 2008.     

Pellet-cladding interactions and length changes in SGHWR fuel pins

Fuel pins in the Steam Generating Heavy Water Reactor increase in length during irradiation, due largely to a mechanical interaction between the cladding and the fuel pellets. The length change produced by this process depends critically on the deformation behaviour of the fuel pellet under the interaction stresses. This paper presents a model of the pellet deformation which has been incorporated into a computer programme, SLIM, capable of predicting extensions from this and other causes in SGHWR fuel pins subjected to any power history. SLIM is shown to give good agreement with measured length changes for all variants of fuel rating, burn-up and pin type tested, and predicts that extension of the narrower pin designs now in use should be reduced by a factor of 0.7 compared with similarly rated standard pins.     

Release and detection of fission products from defective fuel pins

To test the long term behaviour of UO₂-pins with artificial cladding failures a special FR2 inpile steam loop was built. The individual activity concentrations of gaseous and volatile fission products were measured with the aid of a Ge(Li)-system and in addition the integral activity concentration was recorded by a precipitator. Typical results and time behaviours are given. With the aid of a prototype-DND-monitor for the SNR the delayed neutrons of the fission products are measured. Considerable deviations hitherto unknown were observed between the measured values and the values calculated according to the recoil model. The ratio of the both values, called k factor, was in the range of 0.7–533; in the case of fresh fuel the k factor decreases to one-tenth of the initial value after one or two weeks; the k factor of fuel with a burn-up of 46 000 MWd/t was only 1/30 compared to unirradiated fuel and about 1.In the meantime, also in the case of sodium cooling, k factors > 1 were observed, and the GfK defect pin program will focus attention on this effect. The release values of the gaseous fission products were about the same as the values for the neutron precursors, Br and J.     

Role of FP iodine in irradiation behaviors of nuclear fuel pins

During the operation of a reactor, the fuel ($\text{(U,Pu)O}{}_{\mathrm{2-x}}$ or $\text{UO}{}_{\mathrm{2\; +\; x}}$) reacts with its cladding (stainless steel or zircaloy). The role of iodine, a fission product, in this reaction has been examined. Out-of-pile experiments have been done to study the chemical vapor transport of components of stainless steel or pure iron due to iodine released by the electron bombardment of CsI vapor, with or without the excess cesium vapor, using X-ray microprobe analysis. Though addition of excess cesium vapor diminishes the transport rate of stainless steel components, there is still a possibility that cladding components are transported due to iodine in a reactor. Thermodynamic marker tests have been done to check the iodine pressure in the electron-irradiated CsI atmosphere. A short review of the previous relevant studies is also given.     

Radial distribution of bonded fission gas in mixed carbide fuel pins

The fission gas xenon bonded in bubbles, in pores, and in the lattice of mixed carbide fuels is measured by electron-probe microanalysis. Radial xenon distribution and release curves are determined and are calibrated by gas chromatography of the bonded fission gas and by burnup analysis in the respective pin sections of the irradiation experiments FR2 6A and 6C, Mol 11/K 2, and DFR 330/1. The results are correlated to the microstructure of the fuel, bonding medium, temperature, and burnup.     

Analytical investigations of irradiated inert matrix fuel

Yttria stabilised zirconia (YSZ) inert matrix fuel (IMF) fabricated at PSI and irradiated 3 years in the Halden Material Test Reactor (HBWR) since 2000, has been examined by Electron Probe Microanalysis (EPMA) and Secondary Ion Mass Spectroscopy (SIMS) after irradiation and compared with data gained for the unirradiated material. The examined pellet cross-section was estimated to have an equivalent burn-up of 22 MW d kg⁻¹. EPMA measurements demonstrate that the burn-up was rather flat over more than the half pellet radius. A Pu consumption of about 2.5 wt% has been measured with a higher rate in the fuel border zone. The high fuel temperature is responsible for a certain homogenisation of the mineral phases in the fuel centre region whereas the border zone has remained rather with an as-fabricated phase distribution. The central part was also characterised by a dense porosity distribution as well as a temperature and relocation driven depletion of the volatile fission products Xe and Cs. In addition, SIMS has been realised on the same specimen in order to determine the semi-quantitative distribution of different isotopes in the pellet.     

The effect of thermal diffusion from fuel pellets on rewetting of overheated water reactor pins

This paper presents the results of a finite difference solution of a conduction equation for the rewetting of a hot tube containing a filler material. The results show that the effect of a filler is always to reduce the rewetting velocity compared with an empty tube and reasonable agreement is obtained with previous experimental work. The effects of a gas gap on the rewetting of a UO₂-filled Zircaloy tube are discussed. A simple physical model is also presented which shows that the dominant parameter in determining the effect of a filler is (kpc)^1/2. It is suggested that previous theories for rewetting rate derived for empty tubes can be modified to include the effects of a filler by the use of a conduction correction term.     

Rupture and fragmentation of pressurized pipes and fast reactor fuel pins

A model of axial crack propagation in a pressurized tube is developed which predicts the crack velocity and deformation geometry and the minimum driving pressure. Emphasis is placed upon the stability of propagation. The model also offers a criterion for the appearance of multiple cracks and subsequent fragmentation of the tube wall due to excessive axial bending strains. The model is applied to the rupture of gas pipelines, PWR coolant pipes and fast reactor fuel pins.