Effects of Annealing under High Static Pressure on Precipitation Behavior of Fission Gas Bubbles in Irradiated UO2

Annealing experiments were carried out on irradiated UO₂ in argon gas under high pressure (600 and 1,000 kg/cm²) as well as atmospheric, at temperatures of 1,400°–1,600°C. The effects of high external pressure on the behavior of fission gas bubbles in the irradiated UO₂ were studied by comparing replica electron micrographs of fractured surfaces of specimens annealed under different temperatures and pressures. The results indicate that high pressures such as above 600 kg/cm² can be effective in surpressing the growth of fission gas bubbles in both intergranular and intragranular zones, and in inhibiting the joining together of intergranular bubbles to form direct passages for fission gas release.     

Measurement of 239Pu Resonance Fission Integral

An out-of-reactor experiment was conducted for the purpose of studying the external- pressure creep behavior of zircaloy-2 fuel cladding tubes. The hour-glass shape of UO₂ pellets acquired by thermal distortion was simulated by flanges machined out at both ends of stainless steel pellets 21 mm long. The mock-up specimens thus formed were pressurized externally in a furnace, and the changes brought upon the tube diameter were measured at intervals.

The external-pressure creep deformation was observed to proceed in three steps- diametral decrease, elliptical deformation and final collapse. In the case of hollow tubes tested devoid of the pellets, elliptical deformation was observed, which accelerated with time until abrupt final collapse. Elliptical deformation was not observed on the pellet-filled tubes.

Empirical equations were derived from the experimental results as functions of time, of hoop stress and of temperature, to express the external-pressure creep strain behavior of the stress-relieved tube, pellet-filled and thus internally supported at intervals of 20 mm.

No difference in the external-pressure creep deformation was observed between stress-relieved and recrystallized tubes under the condition of 350°C and 72.5 kg/cm².

The pellet-filled specimens showed larger deformation than the hollow tubes in the process of diameter reduction. A certain length of unsupported distance in the range of 0–20 mm appeared to maximize the external-pressure creep deformation at the pellet center, under conditions similar to that of the present experiment.     

Rod Overpressure/Lift-off Testing at Halden In-pile Data and Analysis

The cladding lift-off experiments at Halden yield direct data for the maximum pressure to which a rod can be operated without causing a lasting fuel temperature increase. UO₂ or MOX fuel segments irradiated to high burnup in light water reactors are equipped with a fuel thermocouple and a cladding extensometer. Gas lines attached to the end plugs are connected to a high pressure system for pressurisation with argon and a low pressure system for hydraulic diameter measurements to study cladding outward deformation and axial gas communication within the fuel rod.

The first experiment of the test series utilised a UO₂ fuel segment irradiated in an LWR to 52 MWd/kgUO₂. The test was operated for 4,400 h PWR conditions (155 bar, 310°C) provided by a loop system. The rod was pressurised starting at 205 bar and increasing to 455 bar in steps of 50 bar, while recording fuel centreline temperature and cladding elongation. The hold times at the different pressure levels were long enough to assess temperature trends.

The measured rates of fuel temperature increase suggest that the necessary overpressure to cause a discernible lasting temperature change was 130–145 bar, equivalent to a cladding hoop stress of 70–77 MPa.     

Dancoff Factor in Arrayed-Type Cladded Fuel Lattice

The initial-stage sintering mechanism of hyperstoichiometric urania prepared by sol-gel process was determined in relation to temperature during constant rate heating (CRH). The urania powder used in this experiment was prepared by crushing in Ar atmosphere the micro- spheric gel of UO₂ obtained by sol-gel process, and reducing the resulting powder by heating in H₂ for 1 hr at 500°C. The results obtained from densification measurements indicated that the initial-stage sintering proceeded in two phases governed by different shrinkage mechanisms, as follows.

1. The sintering up to 675°C would be due to a mechanism such as rearrangement of grains and/or plastic flow.

2. Sintering from 750° to 800°C was interpreted as being controlled by uranium volume diffusion.

The estimated diffusion coefficient D = 1.42×10^?6 exp(-52,500/RT) cm²/sec. This value agreed in order of magnitude with the uranium diffusion coefficients measured by other workers for hyperstoichiometric urania.     

Decrease of Cesium Release from Irradiated UO2 Fuel in Helium Atmosphere under Elevated Pressure of 1.0 MPa at Temperature up to 2,773K

In the case of severe accidents, the radionuclides release from fuel could mostly occur at high temperature under elevated pressure. The effect of temperature on the release has been clarified in many previous studies while the pressure influence has been scarcely investigated so far due to difficulty in the experimental operation. To investigate the effect of pressure on the release, two tests under the same conditions except for the system pressure were performed in the VEGA program at JAERI by heating up the irradiated UO₂fuels up to 2,773 K in inert helium. The test results uniquely showed that the release rate of cesium for the temperatures below 2,773 K at 1.0 MPa could be suppressed by about 30% compared with that at 0.1 MPa. This article describes the outlines of the two tests and the observed effects of system pressure on cesium release as well as the results of various post-irradiation examinations. Moreover, the mechanisms and models that explain the pressure effect are discussed.     

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}     

Improvement of UO2 Pellet Properties by Controlling the Powder Morphology of Recycled U3O8 Powder

Powder morphology evolution of recycled U₃O₈ according to the thermal treatments has been studied. The defective UO₂ pellets are oxidized to U₃O₈ powders at a conventional temperature of 350 or 450°C in air. Those powders are pressed into green pellets and then sintered at 1,500 and 1,730°C in H₂ gas flow. Final reoxidized U₃O₈ powers are obtained by reoxidizing those sintered pellets at 450°C in air. This paper shows that the reoxidized U₃O₈ powder morphology and the BET surface areas are greatly dependent on the density of sintered UO₂ pellets before reoxidation. Reoxidized U₃O₈ powders are added to virgin UO₂ powders to fabricate UO₂ pellets and the effect of such addition on the UO₂ pellet properties is investigated. The reoxidized U₃O₈ powders having a certain range of BET surface area significantly promote the grain growth of UO₂ pellets.     

Thermal Recovery of Radiation Defects and Microstructural Change in Irradiated UO2 Fuels

Abstract

Thermal recovery of radiation defects and microstructural change in UO₂ fuels irradiated under LWR conditions (burnup: 25 and 44 GWd/t) have been studied after annealing at temperature range of 450-1,800°C by X-ray diffractometry and transmission electron microscopy (TEM). The lattice parameter of as-irradiated fuels increase with higher burnup, which was mainly due to the accumulation of fission induced point defects. The lattice parameter for both fuels began to recover around 450-650°C with one stage and was almost completely recovered by annealing at 850°C for 5 h. Based on the recovery of broadening of X-ray reflections and TEM observations, defect clusters of dislocations and small intragranular bubbles began to recover around 1.150–1,450°C. Complete recovery of the defect clusters, however, was not found even after annealing at 1,800°C for 5h. The effect of irradiation temperature on microstructural change of sub-grain structure in high burnup fuels was assessed from the experimental results.     

Radiolysis of Dyes in Aqueous Solutions, (V)(1)–(4)

Abstract

Fission gas behavior of UO₂ fuel pellets with controlled microstructure irradiated to 23 GWd/t in a test reactor has been studied by using a postirradiation annealing experiment. Four types of fuel pellets with or without additives were examined : (1) un-doped standard (grain size: 16/μm), (2) un-doped large grained (43μm), (3) 0.7 wt% Nb₂O₅-doped large grained (110/μm), (4) 0.2wt% TiO₂-doped large grained (85μm) fuels. The annealing was conducted at 1,600 or 1,800°C for 5 h in reducing or oxidizing atmospheres.

Fission gas release and bubble swelling caused by the high temperature annealing for the two un-doped fuels were reduced to about 1/3–1/2 by increasing the grain size from 16 to 43 μm, which roughly corresponded to the ratio of their grain sizes. By contrast, the performance of the two large grained fuels doped with Nb₂O₅ or TiO₂ was roughly equivalent to, or rather inferior to that of the standard fuel, despite their large grain sizes of 110 and 85 μm. The fission gas behavior of un-doped fuels was aggravated by increasing the oxygen potential in the annealing atmosphere, while that of additive doped fuels did not depend on it. The effects of grain size, additive doping and oxygen potential on fission gas release and bubble swelling were discussed in connection with the diffusivities of fission gas atoms and cation vacancies.     

Thermal Conductivity Measurements on UO2+x from 300 to 1,400 K

In order to evaluate the thermal conductivity of oxidized fuel pellets of leaker fuel rods, UO_2+x samples with x between 0.00 and 0.20 were prepared by an oxidation or a sintering method. Sample thermal diffusivities were measured by using a laser flash method from 300 to 1,400 K and their thermal conductivities were evaluated from multiplying the thermal diffusivities by the sample densities and the specific heat capacities derived from the literature. The thermal conductivities of UO_2+x were decreased with increasing hyperstoichiometry and they were expressed as a function of their hyperstoichiometry using the concentration of U⁵⁺ formed with the excess interstitial oxygen atoms.     

Effects of UO2/Zircaloy-4 Mole Ratios on Reaction of UO2 Fuel with Molten Zircaloy-4

Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2,223-2,373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2. The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio. Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the U0₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.     

Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.     

Change of Composition of UC1-xNx in Nitrogen Gas

The UC-UN solid solution, UC_1-xN_x, exists stably in monophase within a certain range of nitrogen pressure at constant temperature. At the upper limit of this range, UC_1-xN_x, coexists with UC₂, or with carbon, and it precipitates metallic uranium at the lower limit which corresponds to the decomposition pressure of UC_1-xN_x.

In order to measure the decomposition temperature of UC_1-xN_x with given x under constant nitrogen pressure, it is necessary to heat UC_1-xN_x without changing its composition, at a constant pressure.

In the present work, preliminary considerations have been given to change in composition. Experiments were also performed in which the solid solution was heated at a rate of 200°C/min under different nitrogen pressures.

From the results, it is concluded that there exists a temperature range within which the value of x in UC_1-xN_x, is maintained constant when the nitrogen pressure is fixed.     

Numerical Analysis of Pressure Effects on Heavy Water Separation Characteristics for a Pair of Dual Temperature Multistage-Type H2/H2O-Exchange Columns

In order to study how the operating pressure has influence on the heavy water separation characteristics for a pair of dual temperature multistage-type H₂/H₂O-exchange columns, the authors utilized the analytical expressions derived from the finite difference equation describing the material balance in the exchange column. The separation characteristics of the system were studied theoretically under the conditions where the temperature in a hot column t_h=200°C, the pressure in a hot column π≦50atm and the pressure in a cold column π = 5–50atm.

As the results, the optimum modified specific column volume v^*_opt decreases with the increase of π and then keeps the nearly constant value, and is smaller at ≦=50atm than at ≦=30atm in the case where the temperature in a cold column t_c is constant. Other important parameters such as the optimum operating temperature in a cold column t_copt the optimum numbers of stages in a hot column M_opt and in a cold column t_opt, the optimum hydrogen to feed water molar flow ratio γ _opt and the optimum degradation ratio δ_opt were also calculated.     

Numerical Calculation of H2-HT Separative Performances up to −0.3 MPa of Thermal Diffusion Column with Fairly Higher Hot-Wire Temperature

By using the two-dimensional rigorous numerical solution of flow and convection-diffusion equations, the H₂-HT separative performances of thermal diffusion column with 15 mm-radius and 288.15K cold-wall was analyzed up to ?0.3 MPa for higher hot-wire temperature (up to ?1,700K). Flow analysis has revealed: (1) The magnitude of the free convection is almost proportional to the pressure, and laminar solution of free convection could not be obtained at the pressure more than ?0.32 MPa. (2) The magnitude of the free convection increases gradually with ΔT (the temperature difference between hot and cold surfaces), when ΔT<–800 K. In the range of larger ΔT, the magnitude is almost constant or rather decreases gradually with ΔT. As a result, the laminar solution could always be obtained at the pressure less than ?0.32 MPa, no matter how large ΔT may be. Separative analysis for H₂-HT isotope separation has made clear that the thermal diffusion column with 288.15 K cold-wall should be operated at (1) 0.15–0.2 MPa, (2) ΔT that is as large as technically possible, and (3) the feed rate F of 50–100 cm³ (288.15 K, 0.1 MPa)/min.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

A mechanism for the sintered density decrease of UO2–Gd2O3 pellets under an oxidizing atmosphere

A mixture of UO₂ and Gd₂O₃ powders was pressed into compacts and sintered under various atmospheres ranging from reducing to oxidizing gases. The sintered density of UO₂–10 wt% Gd₂O₃ pellets decreases with increasing oxygen potential of the sintering atmosphere. Dilatometry and X-ray diffraction studies indicate that the delay of densification takes place between 1300°C and 1500°C, along with the formation of (U,Gd) O₂. A very large solubility of Gd₂O₃ in UO₂ relative to the reverse solubility might cause Gd ions to diffuse into UO₂ so directionally that new pores are produced at the places of Gd₂O₃ particles. The new pores may be difficult to shrink and thus lead to the density decrease under an oxidizing atmosphere but not under a reducing atmosphere, because a driving force for the shrinkage of new pores may be smaller under an oxidizing atmosphere than under a reducing atmosphere.     

Densification behaviour of UO2–50%PuO2 pellets by dilatometry

The sintering behaviour of UO₂–50%PuO₂ pellets has been studied using a dilatometer in inert, reducing and oxidising atmospheres. The shrinkage begins at a much lower temperature in oxidising atmosphere such as CO₂ and commercial N₂. The shrinkage rate was found to be maximum for pellets sintered in N₂ atmosphere. The mechanism for the initial stage of sintering was found to be volume diffusion for both oxidising and reducing atmospheres. The activation energy for the initial stages of sintering was found to be 365 and 133 kJ/mol for Ar–8%H₂ and CO₂ atmospheres, respectively. The activation energy obtained using the Dorn method matches well with that obtained using the rate controlled sintering process. The lower activation energy obtained in the oxidising atmosphere is explained with the help of models available in the literature.     

Validity of Constant Collision Density Assumption in Intermediate Energy Region

Several kinds of coated fuel particles, with their coating either intact or artificially cracked, were heated out-of-pile in such manner as to create a sharp temperature gradient across the particles (60°120°C per particle), at temperatures from 1,500° to 1,950°C. The purpose was to obtain information on the displacement of the kernel material relative to the coating. To examine this amoeba effect, the particles were observed, after heating, by both ceramography and ×-ray radiography. The results revealed that:

(1) In the case of UO₂ kernel with artificially impaired coating, their kernels were found to move more readily toward the crack, regardless of the temperature gradient, as compared with UC₂.

(2) The amoeba effect is observed even in out-of-pile heating on intact coated particles with UO₂kernel which moves down the temperature gradient. This UO₂ movement was given a new explanation based on the evaporation and subsequent condensation of the UO₂ within the particle, when the coating is intact.

(3) In case of UC₂ kernel, which moves up the temperature gradient, the sealing-in of the kernel by the intact coatings appears to assume a controlling factor, and the occurrence of evaporation is negligible.     

Sodium Vapor Deposition,on Simulated Model of LMFBR Rotating Shield Plug Annulus, (I)

Experimental study was made to investigate the controlling factors on the vapor deposition rate on reactor operational shield plug annulus, which is exposed to the vapor entrained cover gas during reactor operation. Two simulated test assemblies having annuli were made for this purpose and were installed into a small test vessel. In the experiment, the average deposition rates on the annular walls of the test assemblies were measured for various pool temperatures, and their dependents upon such parameters as pool temperature, T_s (or the saturated vapor pressure P_s at pool surface), cover gas pressure P_g , and temperature drop ΔT_a across cover gas, were studied.

The results revealed that the dominant controlling factor was the vapor pressure P_s at pool surface. Dependent of the average deposition rate φbar_d. upon the above parameters was simply expressed by: φbar_d=Bσp_sD_sΔT_G , where, σ_s is the saturated vapor concentration at pool surface, D_s , the vapor diffusion coefficient, and B the proportional constant.

To these experimental results, the previously published evaporation rate data and the theoretical evaporation rate equation based on Epstein & Rosner's theory were reviewed. Then correlation between the deposition and the evaporation rates was discussed.