Magnus Limbäck

Advanced doped UO2 pellets in LWR applications

The nuclear industry strives to reduce the fuel cycle cost, enhance flexibility and improve the reliability of operation. This can be done by both increasing the fuel weight and optimizing rod internal properties that affect operational margins. Further, there is focus on reducing the consequences of fuel failures. To meet these demands Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO2 fuel containing additions of chromium and aluminium oxides. This paper presents results from the extensive investigation program which covered examinations of doped and reference standard pellets both in the manufactured and irradiated states. The additives facilitate pellet densification during sintering and enlarge the pellet grain size. The final manufactured doped pellets reach about 0.5% higher density within a shorter sintering time and a five fold larger grain size compared with standard UO2 fuel pellets. The physical properties of the pellets, including heat capacity, thermal expansion coefficient, melting temperature, thermal diffusivity, have been investigated and differences between the doped and standard UO2 pellets are small. The in-reactor performance of the ADOPT pellets has been investigated in pool-side and hotcell Post Irradiation Examinations (PIEs), as well as in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced fission gas release, improved PCI performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. Fuel segments have been exposed to ramp tests and enhanced power steady-state operation in the Studsvik R2 reactor after base-irradiation to above 30 MWd/kgU in a commercial BWR. ADOPT reveals up to 50% lower fission gas release than standard UO2 pellets. The fuel degradation behaviour has been studied in two oxidizing tests, a thermal-microbalance test and an erosion test under irradiation. The tests show that ADOPT pellets have a reduced rate of fuel washout, as compared to standard UO2 pellets. Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA-96 Optima2 reloads in 2005.

Self-repairing metal oxides

The oxidation of Cu, Zr, and alloys forming chromia, alumina, and zirconia was studied in a closed reaction chamber in O2 gas near 20 mbar. Information on the position of oxide growth has been gained from the 18O/SIMS technique. Rates of O2 dissociation on metal oxides, Au, and Pt have been evaluated from measurements in labeled O2. The experimental results indicate that hydrogen in the metal substrates induces increased metal-ion transport in internal oxide surfaces during oxidation, which leads to increased oxide growth at the oxide–gas interface. Experiments also show that oxides of rare-earth metals (REM) and Pt catalyze the dissociation of O2. An increased rate of O2 dissociation can lead to increased transport of oxygen ions in the oxides and increased oxide growth at the substrate–oxide interface. A balanced transport of metal and oxygen ions in metal oxides that leads to oxide growth at both the metal–oxide and at the oxide–gas interface is found to be favorable for the formation of protective oxides with good adherence to the metal substrate. Depending on the original proporation of metal–to–oxygen ion transport in the oxide, an addition of hydrogen will increase or decrease the oxidation kinetics. In analogy, an addition of REM will increase or decrease the oxidation kinetics, depending on the original proportion of metal-to-oxygen ion transport.

Toward a Better Understanding of Dimensional Changes in Zircaloy-4: What is the Impact Induced by Hydrides and Oxide Layer?

In pressurized water reactors, new operating conditions (higher burnup, new chemistry, etc.) can have an effect on the dimensional stability of the fuel assembly skeleton. Previous studies have shown that the fuel assembly growth is, among others parameters (free growth, creep), strongly driven by corrosion. Oxide layer and hydrides precipitation could both induce an increase of the dimensional parameters. The scope of the present study is, regardless of irradiation effect, to quantify and to understand the separate effects of hydrogen and oxide layers on the Zircaloy-4 dimensional changes. Experimental works have been performed in laboratory on stress relieved annealed (SRA) and recrystallized (RXA) Zircaloy-4 strips. First, the hydrogen impact on dimensional changes has been studied without the effect of the oxide layer. The measurements were performed at room temperature on strips previously pre-hydrided by the gaseous charging method. The hydrogen content of the samples was between 100 ppm and 2000 ppm. Results indicate a linear correlation between hydrogen content and length variation. RXA material is more affected by the hydrogen effect than the SRA material. Nevertheless, in comparison with data issued from out-of-reactor measurement, the impact of hydrides is not sufficient (with irradiation growth) to explain the post-irradiation examinations (PIE) results. To understand these differences, the oxide layer contribution must be quantified. Second, the impact of the oxide layer was therefore studied on RXA Zircaloy-4 strips. Corrosion tests have been performed in autoclave at 360°C in primary water (2 ppm Li-1000 ppm B–H2) on as-received and pre-hydrided materials. To obtain thicker oxide layer within a shorter duration, samples have been also oxidized in furnace at 415°C. Moreover, as no significant hydriding occurs during oxidation in air, we are able to characterize properly the specific effect of the oxide layer. As for hydrides’ effect, an increase of strain is observed as the oxide thickness becomes thicker. The contributions of hydrides and oxide layer are then discussed with regard to the metallurgical properties of the alloy. Finally, all these results are compared with PIE observations. Free growth, hydride precipitation, and oxide thickness seem to be the three main parameters to explain the dimensional changes in Zircaloy-4 observed in reactor.

Characterization of high burnup fuel for safety related fuel testing

Fuel Assemblies designed and fabricated by Westinghouse Electric Sweden (WSE) to reach high burnup have been operated in the Leibstadt nuclear power plant (KKL) for seven cycles attaining an assembly average burnup above 60 MWd/kgU. The irradiation conditions in KKL featured linear heat generation rates ranging from 250 W/cm early in life down to 100W/cm in the last cycle and normal water chemistry with zinc injection. Selected rods have been extracted at both intermediate and final irradiation stages and hot cell examinations have been carried out. The results show that the fuel is well suited for high burnup applications and rod segments have been provided to the OECD Halden Reactor Program, the OECD Studsvik Cladding Integrity Program and the Japanese ALPS program for dedicated high burnup tests with regard to fission as release and cladding lift-off as well as behavior under power transient, RIA and LOCA.

Gas-Tight Oxides - Reality or Just a Hope

A better understanding of the transport properties of gases in oxides is certainly very important in many applications. In the case of metals, a general protection measure against corrosion implies formation of a dense metal oxide scale. The scale should act as a barrier against gas transport and consequently it needs to be gas-tight. This is often assumed but rarely, if ever, confirmed. Hence there is a need for characterization of micro- and/or meso- pores formed especially during the early oxidation stage of metallic materials. This paper presents a novel and relatively straightforward method for characterization of gas release from an oxide previously equilibrated in a controlled atmosphere. The geometry of the sample is approximated to be a plate. The plate can be self-supporting or constitute a scale on a substrate. A mathematical model for calculation of diffusivity and gas content is given for this geometry. A desorption experiment, involving a mass spectrometer placed in ultra high vacuum, can be used to determine diffusivity and amount of gas released with aid of the mathematical model. The method is validated in measurements of diffusivity and solubility of He in quartz and applied in characterization of two Zroxides and one Fe oxide. From the outgassed amounts of water and nitrogen the H2O/N2 molar ratio can be used to estimate an effective pore size in oxides.

Study on Incipient Cracks at Inner Surface of Cladding Liner after High Power Irradiation Test

10x10 BWR segmented rods with a burnup of about 30 MWd/kgU were subjected to high power irradiation tests in the R2 reactor. Metallography on the cladding showed small incipient cracks with a depth of 10mm at the inner surface of the cladding liner after the irradiation tests. In order to investigate the incipient cracks, several post irradiation examinations were performed. Local micro-hardness of the liner was measured with an indentation force of 1 g. The results showed that the hardness was significantly increased at the inner surface facing the fuel. Electron probe micro-analysis (EPMA) of the liner gave an elemental profile showing an increased amount of fission products at the inner surface and the profile agreed with the micro-hardness profile. Scanning electron microscopy showed propagation of the incipient cracks along grain boundaries. Elemental mapping with EPMA showed concentrated cadmium at the incipient cracks. Considering these observed facts, the cause of the crack formation could be stress corrosion cracking (SCC). Applied hoop stress history to the liner during the irradiation test was calculated with a fuel performance analysis code. Based on the results of the observation and calculation, the process of the crack formation and propagation was discussed.

Investigation of Irradiation Hardening and Embrittlement of Zr-2.5%Nb Alloy with High-Energy (e,γ)-Beams

The paper researches into the effect of the initial structural condition of Zr-2.5%Nb alloy, which is widely used in reactor engineering to manufacture reactor core components (mainly as a structural material for RBMK and CANDU pressure tubes), and of irradiation conditions (temperature, applied stress) on irradiation hardening and embrittlement. The reactor damage of Zr-2.5%Nb alloy was modeled with a method of high-energy 225 MeV (e,γ)-beam irradiation, which allows samples to be irradiated under strictly controlled stress conditions. The research has been carried out on Zr-2.5%Nb alloy exposed to four types of thermo-mechanical treatment. It has been found that the alloy is susceptible to intensive irradiation embrittlement irrespective of its initial condition, and the intensity of Zr-2.5%Nb irradiation hardening greatly depends on pre-treatment. In contrast to other conditions, Zr-2.5%Nb alloy is virtually not susceptible to irradiation hardening after high-speed high-frequency (SHF) heating, quenching, and subsequent annealing in the high-temperature range of the α-region, during which the double-phase α+βNb state with high dispersion of βNb (∼1023 m−3) precipitates develops. The obtained results are in good agreement with the post-reactor irradiation tensile test results. The study demonstrates the efficiency of high-energy (e,γ)-beam irradiation for investigating the irradiation hardening and embrittlement of zirconium alloys.

Effects of Pt Surface Coverage on Oxidation of Zr and Other Materials

Certain elements, including noble metals, are identified to influence corrosion behavior of many metals in high-temperature water/steam and O-2. We have previously reported effects of porous Pt coa ...