Behaviour of fission gas in the rim region of high burn-up UO2 fuel pellets with particular reference to results from an XRF investigation

XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8–54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.     

MOX Fuel Performance Experiment Specified for Japanese PWR Utilisation in the HBWR

In order to obtain high burn-up MOX fuel irradiation performance data, SBR and MIMAS MOX fuel rods with Pufissile enrichment of about 6 wt% have been irradiated in the HBWR. In-pile performance data of MOX have been obtained, and the peak burn-up of MOX pellet have reached to 66 GWd/tM as of October 2004. MOX fuel temperature is confirmed to have no significant difference compared to UO₂, if taking into account adequately for thermal conductivity degradation due to PuO₂ addition and burn-up development, and measured fuel temperature agrees well with HB-FINE code calculation up to high burn-up region. Fission gas release of MOX is possibly larger than UO₂ based on temperature and pressure assessment. No significant difference is confirmed between SBR and MIMAS MOX on FGR behaviour. MOX fuel swelling rate agrees well with solid swelling rate. Cladding elongation data shows onset of PCMI in high power region. Ramp test data from other experiment programs with various types of MOX fabrication route confirms superior PCI resistance of MOX compared to UO₂, due to enhanced creep rate of MOX. The irradiation is expected to continue until achieving of 70 GWd/tM (MOX pellet peak).     

Corrosion of high burn-up structured UO2 fuel in presence of dissolved H2

The influence of high burn-up structured material on UO₂ corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H₂ for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 × 10⁻¹⁰ M and for Pu at 7 × 10⁻¹¹ M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion.     

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO₂ spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO₂ spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO₂.     

Preparation and characterization of the simulated burnup americium-containing uranium–plutonium mixed oxide fuel

In order to investigate the effect on fuel thermophysical properties when adding americium and selected fission products to uranium–plutonium mixed oxide (MOX) fuel, simulated low decontamination MOX fuel with high burn-ups to 250 GWd/t, has been prepared and subjected to characterization tests, elastic moduli measurements and melting temperature measurement. Elastic moduli for the simulated low decontamination MOX fuel were almost the same level as fuel without americium and fission products and decrease in the moduli was slight with increasing simulated burn-up. The melting temperature of high burn-up, low decontamination MOX fuel may be estimated by using the findings on the effect of americium, plutonium addition and fission products accumulation.     

Potential nuclear explosive yield of reactor-grade plutonium using the disassembly theory of early reactor safety analysis

The disassembly theory of reactor safety analysis developed for early metal-fueled criticals is applied to determining the potential nuclear explosive yield of reactor-grade plutonium. After verification of the theoretical models, materials data, and equation of state by recalculation of published data, this disassembly theory is applied to so-called hypothetical nuclear explosive devices (HNEDs) based on reactor-grade plutonium. The masses for k_eff = 0.98 and the neutron life times are calculated for such devices by applying neutron transport theory and Monte Carlo codes. Spherical shock compression models describe the density variations as a function of space and time for spherical shock compression of the reactor-grade plutonium sphere with a natural-uranium reflector. Reactivity calculations are performed to determine “Rossi alpha” as a function of time during spherical shock compression. Pre-ignition theory shows that pre-ignition by spontaneous fission neutrons occurs just after prompt criticality is achieved. The chain reaction and power excursion initiated lead to internal pressure buildup which stops the spherical shock wave after it has penetrated between 1.3 cm and 1.8 cm into the plutonium metal sphere. This limits the maximum reactivity input and the potential nuclear explosive yield to 0.12 up to 0.35 kt TNT (equivalent).However, these results do not describe the full reality. In a companion paper, thermal analysis shows such HNEDs to be technically unfeasible for all reactor-grade plutonium from spent LWR fuel with a burn-up of more than 30 GWd/t as long as low technology is used in the spherical implosion lenses of chemical high explosives. More advanced medium technology would raise this burn-up limit to approximately 55 GWd/t.     

Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t

The thermal diffusivity and specific heat of reactor-irradiated UO₂ fuel have been measured. Starting from end-of-life conditions at various burn-ups, measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage were determined. In this context, particular emphasis was given to the behaviour of samples displaying the high burn-up rim structure. Recovery stages could be thoroughly investigated in samples that were irradiated at low burn-ups and/or at high irradiation temperatures. Other samples, in particular those exhibiting the characteristic rim structure, disintegrated at temperatures slightly higher than the irradiation temperature. Finally, from a database of several thousand measurements, an accurate formula for the in-pile thermal conductivity of UO₂ up to 100 GWd t⁻¹ was developed, taking into account all the relevant effects and structural changes induced by reactor burn-up.     

Measurement and calculation of neutron densities in BWR high burn-up fuels

A neutron-scanning device was developed for measuring accurate neutron densities of BWR high burn-up fuels up to 65 GWd tU⁻¹. Characteristic test of this device was done with a ²⁵²Cf source and adopted to measure axial distributions of neutron densities of BWR spent fuels with various enrichments (2.0–3.4%), which had been irradiated up to 60 GWd tU⁻¹ at Fukushima Daini Nuclear Power Station Unit 2(2F-2). We found the measured neutron densities were proportional to about fourth power of the corresponding burn-up values. The neutron densities calculated by the ORIGEN2.1 code and various cross section libraries showed good agreements with the measured ones in profile and absolute value except for BWR-UE file mainly based on ENDF/B-IV. The BS240J32 library based on JENDL3.2 was the best among the investigated libraries.     

Corrosion of irradiated MOX fuel in presence of dissolved H2

The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO₃ solution in presence of dissolved H₂ for 2100 days. The results show that dissolved H₂ in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10⁻¹⁰ and 5 × 10⁻¹¹ M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO₂ matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.     

Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO₂ and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO₂ and MOX as a function of burn-up and irradiation temperature is proposed.     

VALMOX: validation of nuclear data for high burn-up MOX fuels

VALMOX, an acronym for validation of nuclear data for high burn-up MOX fuels, is one of the projects of the cluster evolutionary fuel concepts: high burn-up and MOX fuels (EVOL). It covers 30 months, from October 2001 to March 2004.It considers the evaluation of the actinide inventory of MOX fuel at high burn-up (typically 60 GWd/t) in light water reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file.So far, the major effort was placed on the evaluation of MOX fuel irradiations in pressurised water reactors, and first results will be presented and compared.     

Two-step two-stage fission gas release model

A two-step two-stage model is developed in this study on the basis of the recent theoretical model. This model incorporates a two-step burn-up factor in the two-stage diffusion processes in the grain lattice and at the grain boundary during the fission gas release. In-pile data sets available in FRAPCON-3 code are used to validate the model. Results show that the predictions are in better agreement with the experimental measurements than those of any other models built in the code over the entire burn-up range up to 75 000 MWd/MTU.     

Performance of TRISO particles fueled with plutonium and minor actinides in a PBMR-400 core design

The TRISO particle design of high temperature reactors fueled with plutonium (Pu) and/or minor actinides (MAs) is investigated by calculating the failure fraction of TRISO particles during irradiation. For this purpose, a fuel depletion, neutronics and thermal-hydraulics code system, which delivers the fuel temperature, fast neutron flux and power density profiles, is coupled to an analytical stress analysis code. The latter is being further developed for the calculation of a reliable and realistic failure fraction. The code system has been applied to a PBMR-400 design containing TRISO particles fueled with 1st and 2nd generation plutonium and with a target burn-up of 700 and 600 MWd/kgHM, respectively. It is shown that the pebble-bed type high temperature reactor under consideration is a promising option for burning Pu and MAs if very high burn-ups can be achieved. The TRISO particle failure fraction is also calculated for both Pu and MA fuels, and compared to U-based fuel. It is shown by the present stress analysis code that the Pu-based fuel particles need a better design and this has been achieved for the MA-based fuel, in which helium gas atoms have a significant contribution to the buffer pressure.     

Core concept of compound process fuel cycle

This paper presents fast reactor core concept and its feasibility as a part of newly proposed compound process fuel cycle in which spent fuels of light water reactor are multi-recycled without conventional reprocessing but with only pyrochemical processing, fuel re-fabrication and reloading to the fast reactor core. Results of the core survey analyses in order to find out the feasibility of this concept, taking example for BWR MOX spent fuel of 60 GWd/t burn-up, show that four times recycling of LWR spent fuel with the burn-up of more than 300 GWd/t can be achieved without increasing MA content. Such benefits will be expected in this concept as reduction of fuel cycle cost due to simplified reprocessing procedure, reduction of environmental impacts due to reduced high level waste, efficient utilization of nuclear fuel resources, enhancement of nuclear non-proliferation, and suppression of LWR spent fuel pile-up.     

Extension of the TRANSURANUS burn-up model

The validation range of the model in the TRANSURANUS fuel performance code for calculating the radial power density and burn-up in UO₂ fuel has been extended from 64 MWd/kgHM up to 102 MWd/kgHM, thereby improving also its precision. In addition, the first verification of calculations with post-irradiation examination data is reported for LWR-MOX fuel with a rod average burn-up up to 45 MWd/kgHM. The extension covers the inclusion of new isotopes in order to account for the production of ²³⁸Pu. The corresponding one-group cross-sections used in the equations rely on results obtained with ALEPH, a new Monte Carlo burn-up code. The experimental verification is based on electron probe microanalysis (EPMA) and on secondary ion mass spectrometry (SIMS) as well as radiochemical data of fuel irradiated in commercial power plants. The deviations are quantified in terms of frequency distributions of the relative errors. The relative errors on the burn-up distributions in both fuel types remain below 12%, corresponding to the experimental scatter.     

Oxygen potential of high burn-up fast breeder oxide fuel

The oxygen potential of fast breeder mixed oxide fuel (U_0.8Pu_0.2)O_1.98 irradiated to different burn-ups up to 11 at% has been determined between 900 and 1300 K by measurements of the electromotive force in a galvanic microcell. The oxygen potential increases continuously with burn-up, due to the oxidative nature of fission and due to fission products dissolved in the fuel matrix. Lattice parameter measurements of similar fuel indicate that the fuel with the initial O/M ratio of 1.98 is still substoichiometric even at 7% burn-up. If the lattice parameter measurements are accepted, the increase in due to fission products is larger than assumed so far.     

Effects of lithium burn-up on TBR in DEMO reactor SlimCS

Lithium in a breeding blanket is burned up through neutron nuclear reactions in fusion DEMO reactors. Effects of decrease of solid breeder materials due to lithium burn-up on tritium breeding ratio (TBR) are not systematically calculated in the past. For the SlimCS blanket design, TBR is calculated taking into account the lithium burn-ups by one dimensional Sn radiation transport calculation code ANISN in this study. The ⁶Li burn-ups are 8–79% after 10-year operation. TBR due to ⁶Li decreases to 40% of the initial one in some layer, while it increases in some layers. The TBR integrated over all the blanket decreases to around 96% of the initial one. The study makes it clear that the reduction of the TBR due to the lithium burn-up is not so large.     

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of ⁸⁵Kr have been studied for U0₂ pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography.

No significant difference in the release of ⁸⁵Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of ⁸⁵Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration). A relationship between intergranular bubble concentration N_g per unit area of grain face and average bubble diameter d_g was expressed as N_g∝d_{g 2.1}     

MgO-pyrochlore composite as an inert matrix fuel: Neutronic and thermal characteristics

Inert matrix fuels are an important component of advanced nuclear fuel cycles, as they provide a means of utilizing plutonium and reducing the inventory of ‘minor’ actinides. We examine the neutronic and thermal characteristics of MgO-pyrochlore (A₂B₂O₇: La₂Zr₂O₇, Nd₂Zr₂O₇ and Y₂Sn₂O₇) composites as inert matrix fuels in boiling water reactors. By incorporating plutonium with resonance nuclides, such as Am, Np and Er, in the A-site of pyrochlore, the k_infvs. burn-up curves are shown to be similar to those of UO₂, although the Doppler coefficients are less negative than UO₂. The Pu depletion rates are 88-90% (²³⁹Pu) and 54-58% (total Pu) when the inert matrix fuels experience a burn-up equivalent of 45 GWd/tHM UO₂. Because of the high thermal conductivity of MgO, the center-line temperatures of the MgO-pyrochlore composites at 44.0 kW/m are lower than those of UO₂ pellets. After burn-up, the A-site cation composition is 15-35 at.% lower than that of the B-site cations in pyrochlore (e.g., A_1.84B_2.17O_7.00) due to the fission of Pu in the A-site and the presence of fission product elements in the A- and B-sites of the pyrochlore structure.     

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 UO₂ 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.