Relocation of Fission Products and PU in Irradiated UO2 Pellet

Distribution of fission products (¹⁴⁴Ce, ¹⁰⁶Ru, ¹³⁷Cs, ⁹⁵Zr-⁹⁵Nb and ⁸⁵Kr) and Pu were examined in UO² pellets which were irradiated at high temperature (～3,000°C). Two capsules, each containing six natural uranium UO² pellets were irradiated in GETR at 0.05 and 0.10% burnup respectively. The UO² pellets were completely fused into a single rod, and a big void was formed along the length of the rod as a result of the high temperature. Analysis was made by direct γ-scanning along the length and across sections of the rods, as well as by samples scraped off from the rods. Pu and Ru were found to be distributed fairly regularly, while Cs was markedly concentrated at both ends of the rod where the temperature was lower than the middle part. Zr-Nb was found to be concentrated in the inner portions which were the hottest parts. About 2% of the ⁸⁵Kr remained in a sample of 0.05% burnup, though only a trace amount in that of 0.1% burnup. Distribution of γ-activity along the length of the rods was quite irregular, having wide depleted sections, which might be explained from the exposure to high temperature.     

Separation of Fission Products from Irradiated Uranium Dioxide in Fused Nitrate Systems

The separation of fission products from irradiated UO₂ in fused nitrate systems was studied by the following procedure:

Dissolution of UO₂ in fused NH₄NO₃; adsorption of fission products on glass powder; fused salt chromatography with γ-Al₂O₃; evaporation from fused nitrates; and precipitation of U-compound (probably alkali uranate, M₂U₂O₇) in a fused LiNO₃-KNO₃ mixture. Radiochemically pure Zr-Nb was selectively separated from the fused NH₄NO₃ melt by adsorption. Ru and I were completely evaporated from fused LiNO₃-KNO₃ mixture at above 280°C and from fused NH₄NO₃at above 240°C, respectively. By means of chromatography with a γ-Al₂O₃ column, Zr-Nb and rare earths were completely separated from U in fused NH₄NO₃ or fused NH₄NO₃-LiNO₃ medium at 150° to 180°C, and further, Cs, Ba and Ru were also expected to be separable from U under suitable condition. When the U-compound was precipitated in fused LiNO₃-KNO₃ at 350°C, Cs, Sr, Ba and Ru were fairly well separated from U.     

A Preliminaly Experiment on Radial Distribution of Cs-137, Sr-89-90 and Zr-95 in Sintered Uranium Dioxide Pellet

This paper reports on the determination of the radial distribution of fission products (Cs, Sr and Zr) in neutron irradiated UO₂ pellets. The first part gives the reuslts obtained by gradual dissolution of a pellet. In the second part the fission products distribution has been determined by micro-sampling along the cross-sectional surface of a UO₂ pellet.     

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 UO_2+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 UO_2+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.     

The D-COM blind problem on fission gas release: The predictions of the TRANSURANUS and future codes

The concentration of retained xenon, the percentage of porosity and the UO₂ grain size have been measured as a function of radial position in the base irradiated rod AG11-8 and the transient tested rod AG11-10. In the base irradiation, densification of the fuel took place and slight grain growth occurred at the pellet centre. Gas release was not detected. During the transient test, 15–20% of the xenon inventory was released from the fuel grains. Gas release was accompanied in the central region of the fuel by an increase in the porosity from 4.7 to 6–8%. These findings are compared with the predictions made by the fuel performance code TRANSURANUS. The code predictions are in good agreement with the experimental observations. FUTURE was used to investigate the development of gas bubbles and the mechanisms controlling gas release in the rods during the base irradiation and the transient test. According to FUTURE fission gas will have accumulated on the grain boundaries during the base irradiation. The code indicates that variations in the fuel microstructure resulting from the base irradiation will have caused the level of gas release to vary along the fuel stack in rods AG11-9 and AG11-10 during the transient test. FUTURE also suggests that fission induced bubble re-solution became increasingly important for release during the latter stages of the transient test. Moreover, the code calculations imply that bubble migration could have played a significant role in the release process.     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

The radial distribution of fission gases and other fission products in irradiated PWR fuel

The radial distribution of fission gas (xenon) and other fission products (cesium, ruthenium, cerium) has been measured on UO₂ fuel pellets irradiated in commercial pressurized water reactors to burnups between 13.23 and 48.26 GWd/tU. Fission gas release occurs from the pellet center, and at temperatures $\text{< 1300° C}$ is confined to the region of grain growth. The maximum fractional release measured at the center ranges from 20% to 30%. Only at high burnup (48.26 GWd/tU) an additional release of cesium has been observed. This is considered as evidence for an increase in fission product release at higher burnups. At fuel center line temperature $\text{> 1500° C}$ a high fission gas release is accompanied by a high cesium release. The local release starts at the onset of fission gas bubbles precipitating on grain boundaries and saturates in the center of the pellet at a fractional release value of about 90%.     

Investigation on high temperature vapor pressure of UO2 containing simulated fission-product elements

During the hypothetical core disruptive accident (HCDA) of a fast breeder reactor (FBR), the temperature of the fuel would rise above 3000 K. The experimental data concerning the saturated fuel vapor pressure are necessary for the analysis of the HCDA. In this study, the UO₂ containing Cs, Ba, Ag, or Sn was used to simulate the irradiated fuel in the FBR.The saturated vapor pressure of pure UO₂ and UO₂ containing Cs, Ba, Ag, or Sn at 3000 to 5000 K was measured dynamically with a pulse laser and a torsion pendulum. The surface of a specimen on the pendulum was heated to eject vapor by the injection of a giant pulse ruby laser beam. The pressure of the ejected vapor was measured by both the maximum rotation angle of the pendulum and the duration of vapor ejection. The saturated vapor pressure was theoretically calculated by using the ejected vapor pressure. The surface temperature of the specimen was estimated from the irradiated energy density measured with a laser energy meter.The saturated vapor pressure of UO₂ at 3640 to 5880 K measured in this study was near the extrapolated value of Ackermann's low temperature data. The vapor pressure of UO₂ containing Cs, Ba, Ag or Sn was higher than that of UO₂. The saturated vapor pressure of UO₂ and a solid fission products system was calculated by using these experimental data.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

Releases of Cesium and Poorly Volatile Elements from UO2 and MOX Fuels under Severe Accident Conditions

Radionuclide release from fuel under severe accident conditions has been investigated in the VEGA program at the Japan Atomic Energy Agency. In this program, three types of fuel, two UO₂ fuels irradiated at PWR and BWR and a MOX fuel irradiated at the ATR Fugen, were heated up to about 3130K in helium atmosphere at 0.1 MPa. Comparison of experimental data and evaluation with computer code analyses showed that Cs release is essentially identical among the three fuels. The Cs release from fuel may differ below about 1770K due to a difference in migration to grain boundaries during irradiation. The difference was not also observed for releases of poorly volatile elements, namely, U, Pu, Sr and Mo between UO₂ and MOX fuels. The release rate of Pu became slightly higher than that of U at 3130 K. The release rate of Sr increased at 3130 K, while that of Mo was quite low at temperatures above 2310 K.     

Spent fuel radionuclide source-term model for assessing spent fuel performance in geological disposal. Part I: Assessment of the instant release fraction

A source-term model for the short-term release of radionuclides from spent nuclear fuel (SNF) has been developed. It provides quantitative estimates of the fraction of various radionuclides that are expected to be released rapidly (the instant release fraction, or IRF) when water contacts the UO₂ or MOX fuel after container breaching in a geological repository. The estimates are based on correlation of leaching data for radionuclides with fuel burnup and fission gas release. Extrapolation of the data to higher fuel burnup values is based on examination of data on fuel restructuring, such as rim development, and on fission gas release data, which permits bounding IRF values to be estimated assuming that radionuclide releases will be less than fission gas release. The consideration of long-term solid-state changes influencing the IRF prior to canister breaching is addressed by evaluating alpha self-irradiation enhanced diffusion, which may gradually increase the accumulation of fission products at grain boundaries.     

The effect of open porosity on the re-solution of fission gas from grain-edge bubbles

The model of fission gas release from UO₂ fuel during irradiation has been modified to include the effects of the presence of grain-edge porosity on the overall level of re-solution of gas atoms from grain boundaries. The consequence of this improvement is to enhance the predicted fractional gas release at high burn-up.     

Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO₂ fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor k_inf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO₂ fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO₂, UO₂ with 4.0 vol.% BeO, and UO₂ with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO₂ fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the k_inf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO₂.     

Compression and Decompression Waves in Steam-Water System

The ratio of the γactivities per fission from fission products of ²³⁹Pu and ²³⁵U, and its time dependence were measured by double fission chamber technique. The γ-activity from the fission products of ²³⁹Pu fission was lower than the corresponding activity relevant to ²³⁵U fissions. The ratio varied with the cooling time allowed after irradition.

This ratio was applied to power distribution measurements by γ-scanning method in multi-region cores composed of PuO₂-UO₂ and UO₂ fuels. To obtain the relative power, the measured γ-activities from the fission products in the fuel rods were corrected for the difference between the γ-activities per fission from the fission products.     

Multiphysics coupled modeling of light water reactor fuel performance

A fuel performance code for light water reactors called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed. The CAMPUS code considers heat generation and conduction, oxygen diffusion, thermal expansion, elastic strain, densification, fission product swelling, grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, fuel thermal and irradiation creep, cladding thermal and irradiation creep and oxidation. All the equations are implemented into the COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet with cladding. Comparisons of critical fuel performance parameters for UO₂ fuel using CAMPUS are similar to those obtained from BISON, ABAQUS and FRAPCON. Additional comparisons of beryllium doped fuel (UO₂-10%volBeO) with silicon carbide, instead of Zircaloy as cladding, also indicate good agreement. The capabilities of the CAMPUS code were further demonstrated by simulating the performance of oxide (UO₂), composite (UO₂-10%volBeO), silicide (U₃Si₂) and mixed oxide ((Th_0.9,U_0.1)O₂) fuel types under normal operation conditions. Compared to UO₂, it was found that the UO₂-10%volBeO fuel experiences lower temperatures and fission gas release while producing similar cladding strain. The U₃Si₂ fuel has the earliest gap closure and induces the highest cladding hoop stress. Finally, the (Th_0.9,U_0.1)O₂ fuel is predicted to produce the lowest fission gas release and a lower fuel centerline temperature when compared with the UO₂ fuel. These tests demonstrate that CAMPUS (using the COMSOL platform) is a practical tool for modeling LWR fuel performance.     

Effect of Irradiation on the Microstructure of Swaged UO2

A study was made of the microstructure of swaged UO₂, irradiated to a burnup of 3,000 MWD/T-U at an estimated center-line temperature of 1,400°C.

In the peripheral zones of the specimen near the cladding, evidence was observed of sintering of the UO₂ particles by irradiation even at comparatively low temperatures (about 300°–400°C). Tracks of fission product fragments were observed at the inner wall surface of cracks in the specimen. The width of the tracks is seen to be larger than observed on thin films and on sintered UO₂. Mixed phases of UO₂ and fission products precipitated in the interior of crystal grains, observed in the form of hillocks left unetched, were surmounted by several sprouts of substances still more chemically stable. Mixed phases of UO₂ and fission products were precipitated on the inner wall surface of the cracks and appeared upon etching as stepped protuberances of irregular contours. A comparison between the microstructures of swaged UO₂ and sintered UO₂ irradiated under the similar conditions is discussed.     

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.     

Experimental and thermodynamic evaluation of the melting behavior of irradiated oxide fuels

Onset of melting is an important performance limit for irradiated UO₂ and UO₂-based nuclear reactor fuels. Melting (solidus) temperatures are reasonably well known for starting fuel materials such as UO₂ and (U,Pu)O₂, however the influence of burnup on oxide fuel melting behavior continues to represent an area of considerable uncertainty. In this paper we report the results of a variety of melting temperature measurements on pseudo-binary fuel-fissia mixtures such as UO₂-PuO₂, UO₂-CeO₂, UO₂-BaO, UO₂-SrO, UO₂-BaZrO₃, and UO₂-SrZrO₃. These measurements were performed using the thermal arrest technique on tungsten-encapsulated specimens. Several low melting eutectics, the existence of which had previously been inferred from post-irradiation examinations of high burnup mixed oxide fuels, were characterized in the course of the investigation. Also, an assessment of melting temperature changes in irradiated oxide fuels due to the production and incorporation of soluble oxidic fission products was performed by application of solution theory to the available pseudo-binary phase diagram data. The results of this assessment suggest that depression of oxide fuel solidus temperatures by dissolved fission products is substantially less than that indicated by earlier experimental studies.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Dedication

A theoretical model describing the release of unstable gaseous and volatile fission products from irradiated UO₂ fuel is developed. The extent of interconnectivity of grain-edge tunnel pores is determined in order to assess the likelihood of an unstable atom reaching the exterior of the fuel before decaying. The model behaviour is compatible with experimental observation, and the effects of grain size, restraint pressure and temperature within the fuel are discussed.