High Burnup UO2 Research Articles

The effects of irradiation condition and microstructural change on lattice parameter, crystal lattice strain and crystallite size in high burnup UO 2 pellet

Pellet samples (average burnup: 37–62 GWd/t) were prepared from three kinds of fuel rods which were irradiated in the Halden Heavy Water Reactor in Norway, and microstructural changes in the pellet samples were investigated by means of optical microscopy, SEM/EPMA and micro-X-ray diffractometry. The measured lattice parameters tended to be smaller than the values reported previously, and it is likely that the measured lattice parameters were affected by the temperature conditions during the irradiation tests and the microstructural changes which occurred in the samples. Considering the burnup dependence of uniform and non-uniform strains, the following things are suggested: the interstitial atoms which cause uniform strain begin firstly to form dislocations as a recovery process of irradiation defect and the dislocation density increases. With increasing burnup, the accumulation of dislocations in the crystallite saturates and the migration of dislocations becomes dominant as a recovery process of irradiation defects in the crystallite.

Model for evolution of crystal defects in UO 2 under irradiation up to high burn-ups

The model for dislocations evolution under irradiation conditions in UO 2 is developed and implemented in the MFPR code. Being combined with the MFPR set of microscopic equations for the evolution of point defects and their interactions with gas bubbles, a self-consistent consideration of the whole system of point and extended defects in irradiated fuel, including point defects (vacancies, interstitials and gas atoms), as well as extended defects (bubbles, dislocations, vacancy loops and pores), is attained. The MFPR code with the new defect evolution model is successfully validated against steady-irradiation experiments, in which the dislocation density and the bubble concentration and mean size were directly measured as functions of burn-up at ≈1000 K. Being applied to higher temperatures, the code allows mechanistic interpretation of the temperature threshold for the fuel restructuring observed in the rim-zone of high burn-up UO 2 fuel.

SIMS Analysis of Xe and Kr in a High Burn-up UO2 Nuclear Fuel

Extended abstract of a paper presented at Microscopy and Microanalysis 2008 in Albuquerque, New Mexico, USA, August 3 – August 7, 2008

Isotopic Analysis of Actinides and Fission Products in LWR High-Burnup UO2 Spent Fuels and its Comparison with Nuclide Composition Calculated Using JENDL, ENDF/B, JEF and JEFF

A chemical isotopic analysis of the actinides and fission products of a high-burnup PWR-UO2 fuel with an average burnup of 60.2 MWd/kgHM was carried out to accumulate extensive nuclide composition data. Furthermore, computational analysis was performed using the integrated burnup calculation code SWAT. The differences between the amounts obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3, ENDF/B-VI.5, ENDF/B-VI.8, JEF-2.2 and JEFF-3.0 were evaluated as the ratios of the calculated values to the experimental values (C/E ratios). For actinides, the calculated 244Cm amount, which is an important nuclide as a major neutron source in spent fuels, was underestimated. The main sensitive path for 244Cm was therefore investigated by a simple depletion calculation for actinides and the cause of the underestimation of the calculated 244Cm amount is discussed. The fission products 88Sr, 90Sr, 89Y, 106Ru, 133Cs, 135Cs and 144Nd, for which the C/E ratios were larger than 1.05 or smaller than 0.95, were investigated to improve their C/E ratios by a simple depletion calculation with simplified burnup chains and sensitivity coefficients, and the correction values for the fission yields or capture cross sections were estimated. The effect of power history on nuclide composition was also investigated. Additionally, the fission products for which the C/E ratios strongly depend on the type of library are discussed using sensitivity coefficients.

Study on effects of swift heavy ion irradiation in cerium dioxide using synchrotron radiation X-ray absorption spectroscopy

In order to simulate the effects of high energy fission products on high-burnup UO2 nuclear fuel pellets, CeO2 thin films and bulk specimens were irradiated with 200MeV Xe ions. Effects of the irradiation were studied by using Extended X-ray Absorption Fine Structure (EXAFS) measurement and X-ray photoelectron spectroscopy (XPS) at synchrotron radiation facilities. The EXAFS spectra for the irradiated thin films near the Ce K-edge show the formation of the oxygen deficiency around Ce ions. XPS Spectra show that the valence state of Ce atoms is changed from intrinsic Ce4+ to Ce3+ state by the irradiation. As such irradiation effects appear even for low fluence irradiation, the experimental result suggests that the oxygen deficiency and the change in Ce valence state are due to high-density electronic excitation induced by the 200MeV Xe ions.

Measurements of Crystal Lattice Strain and Crystallite Size in Irradiated UO2 Pellet by X-ray Diffractometry

Lattice parameters, crystallite sizes, and nonuniform strains of high-burnup UO2 fuel samples were measured using micro-X-ray diffractometry in order to investigate the effects of grain subdivision and strain distribution between crystallites on the microstructural changes, so-called rim structure formation, in UO2 pellets. The samples were prepared from fuel rods irradiated in commercial reactors, and pelletaveraged burnups of the samples were in the range from 41 to 61GWd/t. While the lattice parameters of the samples increased linearly in the burnup region up to approximately 70 GWd/t, the lattice parameters slightly decreased and tended to level off above 70GWd/t. The measured crystallite sizes were in the range from 100 to 200 nm and these were nearly the same as those of the “recrystalized grains” in the rim structure. The elastic strain energy densities, which were evaluated from the lattice parameters and nonuniform strains, tended to increase with burnup and show two plateaus at burnups of approximately 50 and 70GWd/t, respectively. The saturation of the measured strain energy density at approximately 50GWd/t can be attributed to the tangle of accumulated dislocations, and the increase in the strain energy density above 60GWd/t can be explained by the strain caused by crystallite rotation under the restraint conditions between crystallites.

Measurements of Crystal Lattice Strain and Crystallite Size in Irradiated UO2 Pellet by X-ray Diffractometry

Lattice parameters, crystallite sizes, and nonuniform strains of high-burnup UO2 fuel samples were measured using micro-X-ray diffractometry in order to investigate the effects of grain subdivision and strain distribution between crystallites on the microstructural changes, so-called rim structure formation, in UO2 pellets. The samples were prepared from fuel rods irradiated in commercial reactors, and pelletaveraged burnups of the samples were in the range from 41 to 61GWd/t. While the lattice parameters of the samples increased linearly in the burnup region up to approximately 70 GWd/t, the lattice parameters slightly decreased and tended to level off above 70GWd/t. The measured crystallite sizes were in the range from 100 to 200 nm and these were nearly the same as those of the “recrystalized grains” in the rim structure. The elastic strain energy densities, which were evaluated from the lattice parameters and nonuniform strains, tended to increase with burnup and show two plateaus at burnups of approximately 50 and 70GWd/t, respectively. The saturation of the measured strain energy density at approximately 50GWd/t can be attributed to the tangle of accumulated dislocations, and the increase in the strain energy density above 60GWd/t can be explained by the strain caused by crystallite rotation under the restraint conditions between crystallites.

Isotopic Analysis of Actinides and Fission Products in LWR High-Burnup UO2 Spent Fuels and its Comparison with Nuclide Composition Calculated Using JENDL, ENDF/B, JEF and JEFF

Isotopic Analysis of Actinides and Fission Products in LWR High-Burnup UO2 Spent Fuels and its Comparison with Nuclide Composition Calculated Using JENDL, ENDF/B, JEF and JEFF

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO2 fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO2 to UO2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.

Summary and Interpretation of the CABRI REP-Na Program

The CABRI REP-Na program was performed in the sodium loop of the CABRI reactor by the French Institut de Radioprotection et de Sûreté Nucléaire. The objective was to study the behavior of high-burnup UO2 and mixed-oxide (MOX) fuel during a reactivity-initiated accident (RIA) and involved eight tests with UO2 and four with MOX fuel. Failures of some UO2 and MOX fuel rods at enthalpy levels ranging from 125 to 472 J/g (30 to 113 cal/g) demonstrated the need for further development of the present safety criteria pertaining to fuel behavior. Detailed interpretation of the test data led to identifying the deleterious influence of a high clad corrosion level with hydride concentrations on clad failure, the contribution of grain boundary gases on fission gas release, and potential clad loading, mainly in MOX fuel during the first phase of the transient without significant clad temperature increase.Questions still remain concerning the transient fission gas behavior and its impact on clad loading during the entire transient, the rod behavior with high clad temperature and internal pressure, and the postfailure phenomena (fuel ejection, fuel/coolant interaction with finely fragmented solid fuel). These issues will be addressed by the CABRI International Program tests under typical pressurized water reactor conditions.

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO2 fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO2 to UO2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.

Estimation of the Grain Boundary Gas Inventory in MIMAS/AUC MOX Fuel and Consistency with REP-Na Test Results

In the frame of its research activities on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” performed the REP-Na program in the CABRI reactor devoted to the study of Reactivity Initiated Accidents. Focused on high burn-up UO2 and MOX fuel behaviour, twelve tests (8 UO2 and 4 MOX) were realized from 1993 to 2000. In all these tests, the influence of grain boundary gas was evidenced and it appeared necessary to perform some estimation of its inventory in irradiated fuel. Such evaluations are presented for the MOX MIMAS/AUC fuel, based on two different approaches: “experimental” and “theoretical.” The fission gas amount located at the grain boundaries increases with burn-up in correlation with the production, but also with the initial Pu enrichment, as soon as the agglomerates have reached the full restructuring threshold for the High Burn-up Structure. The consistency with the REP-Na test results is checked, showing that a significant cladding deformation is needed, clearly higher than for UO2 fuel in order to release all the grain boundary gas in RIA. Furthermore, to the fission gas effect, adds the helium's occluded in the irradiated fuel whose amount increases with burn-up, Pu enrichment and 241Pu and 241Am initial content.

Fission Gas Release Behavior from High Burnup UO2 Fuels under Rapid Heating Conditions

Fission gas release (FGR) behavior under rapid heating conditions of high burnup UO2 fuels with developed rim structure has been examined using two different out-of-pile heating techniques with no restraint pressure. The burnups of the fuel specimens were 36–86 GWd/tU. The bare fuel specimens were heated up to 600-1,800°C at heating rates of 1.7 to 4,600°C/s. The FGR process strongly depended on fuel burnup (extent of rim structure formation) and heating conditions (heating rate, terminal temperature). At lower heating rates below about 10°C/s, growth and interlinkage of fission gas bubbles controlled FGR only at higher temperatures above the threshold temperatures of 1,250--1,450°C, depending on fuel burnup. At higher heating rates above at least 90°C/s, instantaneous FGRs, which originated from the occurrence of microcracks and fuel fragmentation induced by the overpressurization of rim bubbles, arose at higher temperatures above 700°C, only for the high burnup fuels of 74 and 86 GWd/tU with developed rim structure. Almost no FGR was found for non rim-structured fuels, even at the higher terminal temperatures of 1,500--1,800°C. From these and our previous out-of-pile heating results, a map of heating condition dependent FGR under unrestrained conditions is proposed as a function of heating rate and terminal temperature.

Computational Assessment of Burnup-Dependent Fuel Failure Thresholds for Reactivity Initiated Accidents

Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds associated with clad tube failure and fuel incipient melting are determined with respect to fuel burnup in the range from 30 to 70MWd(kgU) 1. The thresholds for clad tube failure are calculated by means of a strain-based clad failure criterion, which rests upon data from more than 200 out-of-pile mechanical property tests on recrystallized Zircaloy-2 and stress relieved annealed Zircaloy-4 cladding. The enthalpy thresholds for fuel incipient melting are calculated, considering the depression of fuel melting point with increasing burnup as well as burnup-related changes to the pellet radial temperature distribution. The calculated enthalpy thresholds for both clad tube failure and fuel incipient melting drop moderately with increasing burnup, and the thresholds are similar for the pressurised and boiling water reactor accidents under consideration. Extensive sensitivity studies are carried out, in order to quantify uncertainties in the calculated bestestimate thresholds.

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO 2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO 2 fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO 2 fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.

Estimation of the Grain Boundary Gas Inventory in MIMAS/AUC MOX Fuel and Consistency with REP-Na Test Results

In the frame of its research activities on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” performed the REP-Na program in the CABRI reactor devoted to the study of Reactivity Initiated Accidents. Focused on high burn-up UO2 and MOX fuel behaviour, twelve tests (8 UO2 and 4 MOX) were realized from 1993 to 2000. In all these tests, the influence of grain boundary gas was evidenced and it appeared necessary to perform some estimation of its inventory in irradiated fuel. Such evaluations are presented for the MOX MIMAS/AUC fuel, based on two different approaches: “experimental” and “theoretical.” The fission gas amount located at the grain boundaries increases with burn-up in correlation with the production, but also with the initial Pu enrichment, as soon as the agglomerates have reached the full restructuring threshold for the High Burn-up Structure. The consistency with the REP-Na test results is checked, showing that a significant cladding deformation is needed, clearly higher than for UO2 fuel in order to release all the grain boundary gas in RIA. Furthermore, to the fission gas effect, adds the helium's occluded in the irradiated fuel whose amount increases with burn-up, Pu enrichment and 241Pu and 241Am initial content.

Fission Gas Release Behavior from High Burnup UO2 Fuels under Rapid Heating Conditions

Fission gas release (FGR) behavior under rapid heating conditions of high burnup UO2 fuels with developed rim structure has been examined using two different out-of-pile heating techniques with no restraint pressure. The burnups of the fuel specimens were 36–86 GWd/tU. The bare fuel specimens were heated up to 600-1,800°C at heating rates of 1.7 to 4,600°C/s. The FGR process strongly depended on fuel burnup (extent of rim structure formation) and heating conditions (heating rate, terminal temperature). At lower heating rates below about 10°C/s, growth and interlinkage of fission gas bubbles controlled FGR only at higher temperatures above the threshold temperatures of 1,250--1,450°C, depending on fuel burnup. At higher heating rates above at least 90°C/s, instantaneous FGRs, which originated from the occurrence of microcracks and fuel fragmentation induced by the overpressurization of rim bubbles, arose at higher temperatures above 700°C, only for the high burnup fuels of 74 and 86 GWd/tU with developed rim structure. Almost no FGR was found for non rim-structured fuels, even at the higher terminal temperatures of 1,500--1,800°C. From these and our previous out-of-pile heating results, a map of heating condition dependent FGR under unrestrained conditions is proposed as a function of heating rate and terminal temperature.

Computational Assessment of Burnup-Dependent Fuel Failure Thresholds for Reactivity Initiated Accidents

Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds associated with clad tube failure and fuel incipient melting are determined with respect to fuel burnup in the range from 30 to 70MWd(kgU) 1. The thresholds for clad tube failure are calculated by means of a strain-based clad failure criterion, which rests upon data from more than 200 out-of-pile mechanical property tests on recrystallized Zircaloy-2 and stress relieved annealed Zircaloy-4 cladding. The enthalpy thresholds for fuel incipient melting are calculated, considering the depression of fuel melting point with increasing burnup as well as burnup-related changes to the pellet radial temperature distribution. The calculated enthalpy thresholds for both clad tube failure and fuel incipient melting drop moderately with increasing burnup, and the thresholds are similar for the pressurised and boiling water reactor accidents under consideration. Extensive sensitivity studies are carried out, in order to quantify uncertainties in the calculated bestestimate thresholds.

Pyroprocessing of Light Water Reactor Spent Fuels Based on an Electrochemical Reduction Technology

A concept of pyroprocessing light water reactor (LWR) spent fuels based on an electrochemical reduction technology is proposed, and the material balance of the processing of mixed oxide (MOX) or high-burnup uranium oxide (UO2) spent fuel is evaluated. Furthermore, a burnup analysis for metal fuel fast breeder reactors (FBRs) is conducted on low-decontamination materials recovered by pyroprocessing. In the case of processing MOX spent fuel (40 GWd/t), UO2 is separately collected for ~60 wt% of the spent fuel in advance of the electrochemical reduction step, and the product recovered through the rare earth (RE) removal step, which has the composition uranium:plutonium:minor actinides:fission products (FPs) = 76.4:18.4:1.7:3.5, can be applied as an ingredient of FBR metal fuel without a further decontamination process. On the other hand, the electroreduced alloy of high-burnup UO2 spent fuel (48 GWd/t) requires further decontamination of residual FPs by an additional process such as electrorefining even if RE FPs are removed from the alloy because the recovered plutonium (Pu) is accompanied by almost the same amount of FPs in addition to RE. However, the amount of treated materials in the electrorefining step is reduced to ~10 wt% of the total spent fuel owing to the prior UO2 recovery step. These results reveal that the application of electrochemical reduction technology to LWR spent oxide fuel is a promising concept for providing FBR metal fuel by a rationalized process.

Core loss during a severe accident (COLOSS)

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy, (b) simultaneous dissolution of UO 2 and ZrO 2, (c) oxidation of U–O–Zr mixtures, (d) degradation–oxidation of B 4C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Break-throughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U–O–Zr and B 4C–metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R&D activities are summarized in this paper.