Lars Hallstadius

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.

Studies regarding corrosion mechanisms in zirconium alloys

Understanding the key corrosion mechanisms in a light water reactor primary water environment is critical to developing and exploiting improved zirconium alloy fuel cladding. In this paper, we report recent research highlights from a new collaborative research programme involving 3 U.K. universities and 5 partners from the nuclear industry. A major part of our strategy is to use the most advanced analytical tools to characterise the oxide and metal/oxide interface microstructure, residual stresses, as well as the transport properties of the oxide. These techniques include three-dimensional atom probe (3DAP), advanced transmission electron microscopy (TEM), synchrotron X-ray diffraction, Raman spectroscopy, and in situ electro-impedance spectroscopy. Synchrotron X-ray studies have enabled the characterisation of stresses, tetragonal phase fraction, and texture in the oxide as well as the stresses in the metal substrate. It was found that in the thick oxide (here, Optimized-ZIRLO, a trademark of the Westinghouse Electric Company, tested at 415°C in steam) a significant stress profile can be observed, which cannot be explained by metal substrate creep alone but that local delamination of the oxide layers due to crack formation must also play an important role. It was also found that the oxide stresses in the monoclinic and tetragonal phases grown on Zircaloy-4 (autoclave testing at 360°C) first relax during the pre-transition stage. Just before transition, the compressive stress in the monoclinic phase suddenly rises, which is interpreted as indirect evidence of significant tetragonal to monoclinic phase transformation taking place at this stage. TEM studies of pre- and post-transition oxides grown on ZIRLO, a trademark of the Westinghouse Electric Company, have used Fresnel contrast imaging to identify nano-sized pores along the columnar grain boundaries that form a network interconnected once the material goes through transition. The development of porosity during transition was further confirmed by in situ electrochemical impedance spectroscopy (EIS) studies. 3DAP analysis was used to identify a ZrO sub-oxide layer at the metal/oxide interface and to establish its three-dimensional morphology. It was possible to demonstrate that this sub-oxide structure develops with time and changes dramatically around transition. This observation was further confirmed by in situ EIS studies, which also suggest thinning of the sub-oxide/barrier layer around transition. Finally, 3DAP analysis was used to characterise segregation of alloying elements near the metal/oxide interface and to establish that the corroding metal near the interface (in this case ZIRLO) after 100 days at 360°C displays a substantially different chemistry and microstructure compared to the base alloy with Fe segregating to the Zr/ZrO interface.

Detailed Analysis of the Microstructure of the Metal/Oxide Interface Region in Zircaloy-2 after Autoclave Corrosion Testing

Two varieties of Zircaloy-2, with different second phase particle (SPP) size distributions and different corrosion resistance, were oxidized in a steam autoclave. Transmission electron microscopy (TEM) of large thin-foil cross-sections of the oxide and the adjacent metal shows an undulating metal/oxide interface in both materials with a periodicity of slightly less than 1 μm and an amplitude of around 100 nm. The SPPs oxidize slower than the surrounding metal, and the absence of volume increase leads to void and crack formation as the SPPs become embedded in the oxide. On SPP oxidation, iron diffuses out of the particles into the surrounding oxide. A sub-oxide with an oxygen content of approximately 50 at. % and a layer thickness of about 200 nm was observed close to the metal/oxide interface. There is a 200 nm oxygen concentration gradient into the metal, from the level close to the sub-oxide of about 30 at. % down to a few atomic percent. All tin in the matrix is incorporated in the sub-oxide, and no segregation to the metal/oxide interface was found.

Autoclave study of zirconium alloys with and without hydride rim

Abstract Autoclave corrosion experiments were conducted on a number of zirconium alloys in different heat treatment conditions. The alloys tested in the present work were Zircaloy-4, ZIRLO® (ZIRLO is a registered trademark of Westinghouse Electric Company LLC in the USA and may be registered in other countries throughout the world. All rights reserved. Unauthorised use is strictly prohibited.) and two variants of ZIRLO with significantly lower Sn levels, referred to here as A-0·6Sn and A-0·0Sn. Typical corrosion kinetics with a change from pre- to post-initial transition was observed with ZIRLO and Zircaloy-4 displaying the shortest time to the initial transition after 120–140 days of autoclave exposure, followed by A-0·6Sn materials after 140–260 days. A-0·0Sn materials showed no sign of transition even after 360 days although one sample tested to 540 days had gone through transition. Material in the stress relieved condition generally experienced initial transition earlier than the same alloy in the recrystallised condition. Pretransition samples had a universally black oxide layer, which eventually developed grey patches when transition occurred. Practically, all non-hydrogen charged alloys showed a strong trend towards cubic oxide growth rates. Cathodic hydrogen charging was conducted to simulate end of life condition of cladding tubes, forming a hydride rich rim region at the outer surface of the cladding tubes. Hydrogen charged materials generally experienced accelerated corrosion of different degrees with the exception of recrystallised A-0·0Sn and partially recrystallised A-0·6Sn showing no sign of acceleration. It therefore seems that increasing tin levels has a negative impact on autoclave corrosion behaviour for materials with and without a hydride rich rim. In developing advanced alloys for use in cladding, this effect has been balanced against the benefits that Sn is known to provide in-reactor, including robustness in corrosion behaviour and reduced irradiation growth. It was noted that most materials with a hydride rich rim exhibit parabolic corrosion kinetics with decreased initial weight gain but increased overall weight gain.

Toward a Comprehensive Mechanistic Understanding of Hydrogen Uptake in Zirconium Alloys by Combining Atom Probe Analysis With Electronic Structure Calculations

The ability of a zirconium alloy to resist corrosion relies on a compromise between two opposing strategies. Minimizing the hydrogen pickup fraction (HPUF) by invoking metallic electron conduction in the barrier oxide results in rapid parabolic oxide growth. On the other hand, slow sub-parabolic barrier oxide growth, as reflected in rate limiting electron transport, may result in a high HPUF. The objective of the present study is to offer mechanistic insights as to how low concentrations of different alloying elements become decisive for the overall corrosion behavior. Combining atomistic microanalysis with first principles modeling by means of density functional theory, the speciation and redox properties of Fe and Ni towards hydrogen evolution are firstly explored. Complementary atom probe microanalysis at the metal–oxide interface provides evidence for Fe and Ni segregation to grain boundaries in Zircaloy-2 that propagates into the ZrO2 scale. Descriptors for how alloying elements in ZrO2 control electron transport as well as catalytic electron-proton recombination in grain boundaries to form H2 are determined by means of theory. The findings are generalized by further atomistic modeling, and are thus put in the context of early reports from autoclave experiments on HPUFs of zirconium with the alloying elements Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, and Nb. A shunting mechanism which combines inner and outer hydrogen evolution mechanisms is proposed. Properties of the transient zirconium sub-oxide are discussed. A plausible atomistic overall understanding emerges.

Oxidation Mechanism in Zircaloy- 2—The Effect of SPP Size Distribution

The metal/oxide interface region in Zircaloy-2 oxidized in autoclave was studied with transmission electron microscopy (TEM) and atom probe tomography. In addition to waviness on the micrometer scale the metal/oxide interface was found to have irregularities on a finer scale, and metal islands were found especially at metal hills (delayed parts of the oxidation front). The thickness of the sub-oxide layer varies considerably along the interface in the same sample, from 100 to virtually 0 nm. The sub-oxide composition may vary on a very fine scale (down to 5nm), and it can sometimes be a mixture of sub-oxides with different oxygen content. The metal matrix in contact with the sub-oxide is saturated with up to 32 at. % oxygen, and the oxygen diffusion profile in the metal is in approximate agreement with literature data for pure Zr. However, the diffusion length appears to be somewhat larger at interface metal hills than under valleys, probably for both geometrical and stress state reasons. Hydride precipitates, hardly visible in conventional TEM, give a good image contrast when employing high angle annular dark field imaging. A model for the oxidation process is presented, where the creep deformation of the metal close to the interface and the formation of lateral cracks in the oxide are of highest importance. The effect of second phase particle (SPP) size is suggested to be twofold: Small and numerous SPPs give a stronger metal and therefore higher stress in the oxide. Small SPPs also nucleate many more lateral cracks in the oxide, which gives a weaker oxide. Together this leads to formation of large cracks associated with transition in the oxidation rate at an earlier time than for a material with larger and fewer SPPs, and thereby a higher oxidation rate.

Effect of Sn on Corrosion Mechanisms in Advanced Zr-Cladding for Pressurised Water Reactors

The desire to improve the corrosion resistance of Zr cladding material to allow high burnup has resulted in a general trend among fuel manufacturers to develop alloys with reduced levels of Sn. While the detrimental effect of Sn on high temperature aqueous corrosion performance is widely accepted, the reason for it remains unclear. High-Energy synchrotron X-ray diffraction was used to characterise the oxides formed by autoclave exposure on Zr-Sn-Nb alloys with tin concentrations ranging from 0.01 to 0.92 wt.%. The alloys studied included the commercial alloy ZIRLO® and two variants of ZIRLO with significantly lower tin levels, referred to here as A-0.6Sn and A-0.0Sn. The nature of the oxide grown on tube samples from each alloy during autoclave testing at 360°C was investigated by cross-sectional Scanning and Transmission Electron Microscopy (SEM & TEM). Non-destructive synchrotron X-ray diffraction analysis on the oxides revealed that the monoclinic and tetragonal oxide phases display highly compressive in-plane residual stresses with the magnitudes dependent on both phase and alloy. Additional in-situ Synchrotron X-ray diffraction experiments during oxidation at 550°C provided further confirmation of the trends seen for autoclave tested samples and demonstrated the presence of elevated levels of tetragonal phase in the initial stages of oxidation. In-situ and ex-situ measurements demonstrate unambiguously that the amount of tetragonal phase present and, more importantly, the degree of transformation from tetragonal to monoclinic oxide both decrease with decreasing tin levels, suggesting that tin stabilises the tetragonal phase. It is proposed that in Zr-Nb-Sn alloys with low Sn, the tetragonal phase is mainly stabilised by very small grain size and therefore remains stable throughout the corrosion process. By contrast, in alloys with higher tin levels larger, stress stabilised, tetragonal grains can form initially, but then become unstable as the corrosion front progresses inwards and stresses in the existing oxide relax.

RIA Failure of High Burnup Fuel Rod Irradiated in the Leibstadt Reactor: Out-of-Pile Mechanical Simulation and Comparison with Pulse Reactor Tests

A high burnup boiling water reactor fuel rod was subjected to reactivity initiation accident (RIA) tests in a research reactor. Two ramp tests were carried out under almost identical irradiation conditions, resulting in cladding failure in a room temperature test, while no failure was observed in a high temperature test. An adjacent segment of the same rod was used for mechanical expansion-due-to-compression (EDC) testing simulating the pellet-cladding mechanical interaction loadings on the cladding during an in-reactor RIA. The EDC tests show the existence of a transition temperature where an abrupt increase in the specimen hoop strain at failure occurs. Additional tests revealed that the transition temperature depends on hydrogen concentration. A possible effect of the rapid heating, which is a specific condition for an in-reactor RIA compared to the static heating in the EDC tests, was verified in the rapid heating/loading tests on unirradiated hydrided Zircaloy, when both loading and heating are performed simultaneously within 50–80 ms. It was shown that strain to failure is dependent on the instantaneous material temperature and is not affected by the pre-heating history. The results show good consistency between the EDC and in-reactor pulse test data. It is concluded that the EDC test can provide valuable information to predict the in-reactor RIA fuel failure.

Method for Analyzing Fission Gas Release in Fuel Rods Based on Gamma-Ray Measurements of Short-Lived Fission Products

Fission gases are produced as a result of fission reactions in nuclear fuel. Most of these gases remain trapped within the fuel pellets, but some may be released to the fuel rod internal gas volume under certain conditions. This phenomenon of fission gas release is important for fuel performance since the released gases can degrade the thermal properties of the fuel rod fill gas and contribute to increasing fuel rod internal pressure. Various destructive and nondestructive methods are available for determining the amount of fission gas release; however, the current methods are primarily useful for determining the integrated fission gas release fraction, i.e., the amount of fission gas produced in the fuel that has been released to the free rod volume over the entire lifetime of a nuclear fuel rod. In this work, a method is proposed for determining the fission gas release that occurs during short irradiation sequences. The proposed method is based on spectroscopic measurements of gamma rays emitted in the decay of short-lived fission gas isotopes. Determining such sequence-specific fission gas release can be of interest when evaluating the fuel behavior for selected times during irradiation, such as during power ramps. The data obtained in this type of measurement may also be useful for investigating the mechanisms behind fission gas release for fuel at high burnup. The method is demonstrated based on the analysis of experimental gamma-ray spectra previously collected using equipment not dedicated for this purpose; however, the analysis indicates the feasibility of the method. Further evaluation of the method is planned, using dedicated equipment at the Halden Boiling Water Reactor.

Texture Evolution of Zircaloy-2 During Beta-Quenching: Effect of Process Variables

The nuclear industry is interested in developing thermomechanical processes to produce random crystallographic orientation (texture) from cold-rolled Zircaloy-2 sheets used to manufacture boiling water reactor (BWR) channels. Randomized textures are beneficial because they minimize anisotropic irradiation-assisted growth, which in turn reduces bowing and uncontrolled deformation of BWR channels during service. The texture evolution of cold-rolled Zircaloy-2 sheets during the allotropic α→β→α phase transformation was characterized by using synchrotron X-ray diffraction in situ and electron backscatter diffraction. The initial strong rolling texture is weakened only if the α→β→α phase transformation is complete. Plastic deformation and grain growth in the β-phase lead to changes in the β texture and modify the inherited α texture. The global texture evolution is not sensitive to levels of stress that do not cause β plastic deformation. These findings demonstrate that accurate temperature control of the β-quenching process is of utmost importance in order to minimize undesirable irradiation growth of BWR channels during service, and that plastic deformation in the β phase can be employed to modify the inherited α texture. BWR channels with β-quenched textures will exhibit minimum irradiation growth caused by texture.