High Burn-up Structure Research Articles

SIMS analysis of an UO 2 fuel irradiated at low temperature to 65 MWd/kgHM

The results of a wide-ranging post-irradiation study of a PWR nuclear fuel by secondary ion mass spectrometry are presented. The average cross-section burn-up and the radial burn-up profile were determined from the radial distributions of one or more of the stable isotopes of fission product Nd. The fission gas Kr was analysed in-situ in a nuclear fuel for the first time and an investigation of the total fission gas content of the high burn-up structure using depth profiling was performed. It was confirmed that Kr is together with Xe in the pores of the high burn-up structure, and that almost all the fission gas lost from the fuel matrix is contained in the pores. In addition to Xe and Kr, the volatile fission products Te, Cs, I and Rb were also detected in the pores. The radial distributions of the minor actinides in the fuel are also reported. It was found that 237Np, unlike the isotopes of Pu, Am, and Cm, does not increase in concentration at the fuel rim.

DISCUSSION ABOUT HBS TRANSFORMATION IN HIGH BURN-UP FUELS

High burn-up transformation process in low temperature nuclear fuel oxides material was observed in the early sixties in LWR UO 2 fuels, but not studied in depth. Increasing progressively the fuel discharge burn-up in PWR power plants, this material transformation was again observed in 1985 and identified as an important process to be accounted for in the fuel simulations due to its expected consequence on fuel heat transfer and therefore on the fission gas release. Fission gas release was one of the major concerns in PWR fuels, mainly during transient or accidents events. The behaviour of such a material in case of rod failure was also an important aspect to analyse. Therefore several national and international programs were launched during the last 25 years to understand the mechanisms leading to the high burn-up structure formation and to evaluate the physical properties of the final material. A large observations database has been acquired, using the more sophisticated techniques available in hot cells. This large database is discussed in this paper, providing basis to build an engineering-model, which is based on phenomenological description data and information accumulated. In addition this paper has the ambition to construct the best logical model to understand restructuring.

Effect of HBS Structure in Fast Release Fraction of 48 GWd/tU PWR Fuel

AbstractTwo important issues in Performance Assessment exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of Instant or better called Fast release Fraction (FRF) and the effect of High Burn-Up Structure (HBS). Therefore this paper focuses on the effect of HBS in FRF of a PWR irradiated in a commercial reactor with a mean Burn-Up (BU) of 48 GWd/tU. Additionally, determined FRF are compared with previous experiments performed with PWR with a mean BU of 60 GWd/tU, in order to evaluate the effect of BU on FRF.In order to study the HBS contribution, static leaching experiments are performed with two samples prepared from different radial positions, labeled Core, central region and Out, enriched with HBS. Two synthetic leaching solutions, bicarbonate and bentonitic granitic groundwater are used under oxic conditions.The estimated FRF are calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP) taking into account the inventory determined experimentally for each fraction.Higher release is observed for Core sample, which can be an indicative that HBS does not increase the RN release. With the exception of Rb and Cs, which release is higher at lower BU, no clear effect of BU is observed within the studied range.

In situ Characterization of UO2 Microstructure Changes During an Annealing Test in an Environmental Scanning Electron Microscope

AbstractA 1 μg High Burn Up Structure (HBS) fragment was extracted from a UO2 fuel pellet irradiated for 7 cycles in a EDF Pressurised Water Reactor (PWR). In situ examinations were performed with an Environmental Scanning Electron Microscope (ESEM) in order to characterize UO2 microstructure evolution during a temperature ramp up to 1,600K. The results are compared to previously published data on HBS annealing tests performed in a Knudsen cell where observed burst releases are explained as sample cracking during the experimental sequence.

Conservative Width of High-Burnup Structure in Light Water Reactor UO2 Fuel as a Function of Pellet Average Burnup

Based on the high-burnup fuel data available in open literature, a conservative width of high-burnup structure (HBS) in light water reactor UO2 fuel, which can be used for fuel performance and accident analysis or assessment of spent fuel under geological disposal conditions, is proposed as a function of pellet average burnup. For pellet average burnup of 30 to 60 GWd/t U, where the HBS generally increases with burnup because of the accumulation of irradiation damage, a conservative HBS width is given by wHBS = 13.3 (buavg - 30), where wHBS is the HBS width in μm and buavg is the pellet average burnup in GWd/t U. For pellet average burnup of 60 to 75 GWd/t U, where microstructural damage caused by irradiation is partly annealed, a conservative HBS width is expressed by wHBS = 2.02 exp(buavg/11.35). In the case of pellet average burnup above 75 GWd/t U up to at least 100 GWd/t U, the HBS width does not exceed some limiting value of 1.5 mm because high temperature in the central region of the fuel pellet has caused an extensive annealing of irradiation damage. In addition, because of significant fission gas release during irradiation up to high burnup, HBS formation might not have expanded to the pellet region whose temperature was lower than the threshold one. Therefore, for this burnup range, a conservative HBS width is given as wHBS = 1500 μm.

Interpretation of High-Burnup Fuel Annealing Tests

The growth of the porosity in high-burnup fuel is of particular interest when considering the effect of fission gas retention within the high-burnup structure (HBS). A mechanistic model of porosity growth under annealing conditions for light water reactor (LWR) UO2 fuel with typical stereological parameters of the HBS has been developed. The model takes into account both multipore and surface interactions that lead to fission gas release from the HBS porosity. We have applied the model to an HBS annealing experiment and found that reasonable agreement with the experimental gas release curve can be achieved for a vacancy diffusion enthalpy close to that measured experimentally. A comparison of the model behaviours at different pore growth rates is performed. We found that the model exhibits porosity evolution characteristics independent of the growth rate, specifically the existence of a maximum porosity, a continuous gas release, and a decreasing pore number density. A general comparison between the model results and in-pile data indicates that up to 250 MWd/kgU, most of the gas is retained within the HBS.

SHIELDED LASER ABLATION ICP-MS SYSTEM FOR THE CHARACTERIZATION OF HIGH BURNUP FUEL

In modem power reactors, nuclear fuels have recently reached 55,000 MWd/MtU from the initial average burnup of 35,000 MWd/MtU to reduce the fuel cycle cost and waste volume. At such high burnups, a fuel pellet produces fission products proportional to the burnup and creates a typical high burnup structure around the periphery region of the pellet, producing the so called 'rim effect'. This rim region of a highly burnt fuel is known to be ca. in width and is known to affect the fuel integrity. To characterize the local burnup in the rim region, solid sampling in the micro meter region by laser ablation is needed so that the distribution of isotopes can be determined by ICP-MS. For this procedure, special radiation shielding is required for personnel safety. In this study, we installed a radiation shielded laser ablation ICP-MS system, and a performance test of the developed system was conducted to evaluate the safe operation of instruments.

Evidence of two gas release kinetics during the oxidation of an irradiated PWR UO 2 fuel

The ‘ADAGIO’ (French acronym for Discriminating Analysis of Accumulation of Inter-granular and Occluded Gas) experiments are dedicated to the determination of the gas inventory located at inter-granular position thanks to a partial oxidation of an irradiated UO 2 fuel. During these experiments the 85Kr gas release is measured as a function of time. The comparison of 85Kr release kinetic with the fuel oxidation kinetic brought into prominence two different gas release kinetics: the first one is proportional to the oxidation kinetic, the second one has a sigmoid shape and could be linked to the gas release of the so called ‘High Burn-up Structure’.

Growth mechanisms of interstitial loops in α-doped UO2 samples

New experimental size distributions are presented for interstitial-type dislocation loops in (U,Pu)O2 samples after 4 and 7 years of self-irradiation along with a new model. For this model, the work of Hayns has been extended to doped materials and to take into account additional phenomena, namely re-solution and coalescence as applied to gas bubble precipitation. The calculations are compared to the experimental size distributions obtained by means of Transmission Electron Microscopy for two different storage times. The role of re-solution and coalescence is discussed based on this model. This constitutes a basis for modelling the high burn-up structure (HBS) formation which is a consequence of the accumulation of radiation damage in nuclear fuels.

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

AbstractThe so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories.A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution.The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.

Interpretation of High-Burnup Fuel Annealing Tests

The growth of the porosity in high-burnup fuel is of particular interest when considering the effect of fission gas retention within the high-burnup structure (HBS). A mechanistic model of porosity growth under annealing conditions for light water reactor (LWR) UO2 fuel with typical stereologieal parameters of the HBS has been developed. The model takes into account both multipore and surface interactions that lead to fission gas release from the HBS porosity. We have applied the model to an HBS annealing experiment and found that reasonable agreement with the experimental gas release curve can be achieved for a vacancy diffusion enthalpy close to that measured experimentally. A comparison of the model behaviours at different pore growth rates is performed. We found that the model exhibits porosity evolution characteristics independent of the growth rate, specifically the existence of a maximum porosity, a continuous gas release, and a decreasing pore number density. A general comparison between the model results and in-pile data indicates that up to 250MWd/kgU, most of the gas is retained within the HBS. Atomic Energy Society of Japan.

Nanocrystalline ZrO 2 ceramics with idealized macropores

Macroporous nanocrystalline 4 mol% yttria-stabilised zirconia ceramics were fabricated by a novel colloidal processing technique, using a commercially available nano-powder with modified surface and polymethylmethacrylate (PMMA) sacrificial templates as starting materials. Monolithic bodies obtained by gel-casting of nano-powder/PMMA-templates suspensions presented remarkably homogeneous microstructures after sintering characterized by grain size <200 nm, macropore diameter around 1 μm, and closed porosity varying between 10 and 20%. The achieved microstructure highly resembles the high burn-up structure of highly irradiated light water reactor UO 2 fuel, the properties of which are aimed to be assessed via the present model material. The sintering behaviour of bodies with or without macropores was compared. The sintering activation energy for dense, PMMA-free bodies was found to be as low as 157 kJ/mol suggesting grain boundary diffusion mechanism for mass transport. For the porous material, the measured sintering activation energy was even lower (129 kJ/mol).

Microstructural Evolution in Cerium Dioxide Irradiated with Heavy Ions at High Temperature

Cerium dioxide (CeO2) is a well-studied oxide because of its technological applications, such as a major component in the catalysts, stable capacitors and so on. CeO2 with a cubic fluorite structure is also expected as the material for simulating behaviors of uranium dioxide under energetic particle irradiation, which leads high burn-up structure attributable to pressure increase in fuel cladding tubes. In this work, the stability of crystal structure and surface morphology change in CeO2 will be clarified. Polycrystalline samples were irradiated with 300 keV Xe+, 18 MeV I5+, 350 keV O+ and 12 MeV O3+ ions at temperatures from 370 to 1000 K. Microstructural evolutions was investigated with Scanning electron microscopy, Raman Spectroscopy and X-ray diffractometry. Irradiation effects at high temperature were observed as surface etching and resulting bubble formation, however blistering was not detected. Raman spectra and diffraction data indicates diffusion of oxygen vacancies is sufficient at 900 K and above.

Detailed characterisations of high burn-up structures in oxide fuels

In this paper, a series of post-irradiation examination results on high burn-up structures (HBS) are summarised for a wide variety of fuels (various programmes on PWR UO 2 fuels with different claddings and with or without doping, PWR MOX fuels, FBR fuels, etc.). This summary is intended to provide a better understanding of how high burn-up structure is formed and subsequently changes up to very high burn-ups in various irradiation conditions. This paper also focuses on the existence of planar defects prior to the HBS formation and their change with time, as well as on the chemical influences and consequences of the HBS. Two new mechanisms are proposed to take into account these observations.

Evolution of porosity in the high-burnup fuel structure

The evolution of the high-burnup structure (HBS) porosity is investigated. Electron probe microanalysis (EPMA) and scanning electron microscope (SEM) measurements of UO 2 fuel with ≈105 GWd/tHM rod average burnup show the formation of an ultra-high burnup structure with a local burnup of 300 GWd/tHM in the proximity of the fuel–cladding interface. Such structure is characterized by gas pores of sizes up to 15 μm. A large population of pores with 3–5 μm pores is also observed in more inner regions of the HBS. An analysis of the pore size distributions indicates predominance of 3.5 μm and 7.5 μm pores. A simple model accounting for vacancy diffusion kinetics and coalescence is used to interpret the observations: the 3.5 μm pores are obtained by growth of 1 μm pores with an initial overpressurization of 50–70 MPa. The extra-large pores with diameters ≈7–8 μm result by coalescence of the intermediate size pores, assuming that: (1) such pores are only slightly overpressurized and that (2) the coalescence process occurs at constant porosity, as observed experimentally.

MARGARET: An Advanced Mechanistic Model of Fission Gas Behavior in Nuclear Fuel

CEA (Commissariat à l'Energie Atomique) develops several fission gas models with different levels of description (more or less detailed or “mechanistic” descriptions). After a synthesis of the main phenomena which have to be considered in a fission gas model, the MARGARET model is presented. Up to five cavities populations (bubbles nucleated during irradiation or fabrication pores) are taken into account, in addition to the dissolved intra- and intergranular gases. The HBS (High Burnup Structure) formation is also considered. The modeling leads to an ordinary differential equations system (typically of 20 non linear coupled equations), which is solved by the Gear method. The computation time remains reasonably short, which allows validation on the whole set of measurements available, integral ones, as well as local ones.

Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO2 rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.

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.

Calculations on fission gas behaviour in the high burnup structure

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing a one-dimensional, mass balance model and apply it to LWR UO 2 fuel at the moderate temperatures found in the rim region. We examine the quantity of gas remaining in the HBS fuel matrix at steady state and compare it with experimental values. We find that the current model reproduces the 0.2 wt% observed xenon concentration under certain conditions, viz. fast grain boundary diffusion and an effective volume diffusion coefficient. A sensitivity analysis is also conducted for the model parameters, the relative importance for which is not well established a priori.

On the thermal conductivity of UO 2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation.