Plutonium Oxide Powder Research Articles

Pressurization Analysis for Heating of a Screw Top Utility Can Loaded With Plutonium Oxide Powder by a 1273 K Fire

AbstractThe documented safety analysis for the Savannah River National Laboratory (SRNL) evaluates the consequences of a postulated 1273 K fire in a glovebox. The radiological dose consequences for a pressurized release of plutonium oxide powder during such a fire depend on the maximum pressure that is attained inside the oxide storage containers. The oxide storage configuration selected for analysis is can/bag/can, comprised of oxide powder inside an 8.38 × 10−6 m3 stainless steel B vial inside 0.006 kg of polyethylene bagging inside a one-quart screw top utility can of the type commonly used to package solvents or rubber cements. To enable evaluation of the dose consequences, temperature and pressure transients have been calculated for exposure of a typical set of storage containers to the fire. The pressurization analysis credits venting to and from the B vial but does not credit venting or leakage from the can. Due to the low rate of venting from the B vial into the can gas space, the can pressure is nearly independent of the B vial pressure. Calculated maximum pressures are compared to the utility can burst pressure. In lieu of a structural analysis of the utility cans, burst pressures and leakage rates were measured using compressed nitrogen gas. The measured gauge burst pressure was 0.250 ± 0.043 MPa. The measured burst pressures are lower than the calculated maximum pressure due to fire exposure, indicating that the utility cans could burst during exposure to a 1273 K fire.

Methods and concerns in disposal of nuclear waste from storage vials post recovery

Plutonium in the form of oxide is stored in plastic vials and bottles during its analysis for trace metal assay. Due to radiation damage and moisture, small amount of oxide samples were found to be adhering to surface of vials/bottles. This required methods for safe disposal. Methods involving wet chemical and dry cleaning procedures were evaluated for effective removal of sticking α active plutonium oxide powder from these vials for their safe disposal. Dry cleaning method for removal of plutonium activity was found to be effective method for disposal of storage vial, while wet chemical method was found to be more suitable for removal of plutonium from damaged PVC bottles. The recovered Pu was checked for chemical purity for 24 analytes using glove box adopted high resolution inductively coupled plasma atomic emission spectrometry.

Surface area and particle size distribution of plutonium oxides derived from the direct oxidation of Pu metal

The intent of this paper is to interrogate the physical characteristics of PuO2 powders and aggregates produced by direct metal oxidation to examine possible relationships between process parameters and the physical characteristics of the product. The majority of the PuO2 samples considered were generated during the pre-production and production phases of the Advanced Recovery and Integrated Extraction System (ARIES) program at Los Alamos National Laboratory (LANL). Oxide formed passively on Pu metal surfaces at ambient atmospheric conditions is also characterized. Plutonium oxide powders from the ARIES Direct Metal Oxidation (DMO) furnace and muffle furnace process lines are very consistent in terms of their bulk and tapped density, specific surface area (SSA) and particle size distribution (PSD). The SSAs of calcined PuO2 generated in the ARIES DMO-2 and muffle furnaces are between 0.1 and 0.5 m2/g. Oxides from both the DMO-2 and muffle furnaces also had similar tri-modal PSDs with maxima at 0.73–3.16 μm, 10.9–18.5 μm, and 41.3–63.0 μm. The PSDs of the oxide formed passively on Pu metal surfaces at ambient conditions were unimodal, with near-Gaussian distributions and D50 close to 1 μm. The oxide formed passively at ambient conditions also had significantly higher SSAs (6–8 m2/g).A documented process upset that originated from a heater element failure in DMO-2 resulted in significantly lower oxidation temperatures, and produced oxide having a slightly greater SSA, and lower bulk and tapped densities. Macroscopic (by quantitative sieving) and microscopic (by laser diffraction of aqueous suspensions) PSDs of the oxide produced during this interval are characterized by depleted large-diameter particle populations. We speculate that the lack of large particles in the aggregate emerging from DMO-2 during the temperature anomaly reflects a change in the thermal regime during oxidation which favors the generation of finer-grained aggregate.

Ultrasonic decontamination of prototype fast breeder reactor fuel pins

Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%.

Nondestructive assay of plutonium in empty stainless steel boxes by apparent mass method

Apparent mass method (Venkataraman and Croft, Nucl Instrum Methods Phys Res A 505:527, 2003), initially standardized for the assay of Pu (Agarwal et al., J Nucl Mater 651:386, 2007) has been used to get Pu amount in empty stainless steel boxes generally used for storing and transferring plutonium oxide powders. The results have been compared with the neutron coincidence counting results and have been found to match well. The advantage of the method is that it can be used for any sample with nonstandard geometry and with uncertain source distribution.

Combined reprocessing of weapons plutonium into high-purity oxide fuel

The main results of a study of regimes for dissolution of plutonium oxide and mononitride powders, obtained by the pyrochemical method from weapons plutonium, in acids for subsequent extraction removal of gallium, americium, and ballast impurities and obtaining ceramic-type plutonium oxide powder suitable for fabricating mixing oxide fuel are presented. It is established experimentally that plutonium oxide and mononitride obtained by the pyrochemical method dissolve rapidly in the acid mixture HNO 3 12 moles/liter and HF 0.1 moles/liter. Plutonium extraction into solution reaches >99%.

Pyrochemical method of salvaging weapons plutonium in oxide for fabricating mixed fuel for fast reactors

Experimental results of investigations of pyrochemical conversion of weapons plutonium into plutonium oxide for fabricating fast-reactor fuel are presented. Weapons plutonium was hydrogenized by hydrogen at 220°C, after which the plutonium hydride obtained was acidified at 550–560°C with the formation of PuO2. To increase fire and explosion safety of the process, a mixture of oxygen with nitrogen, helium, or argon was used or nitriding with nitrogen and oxidation of plutonium mononitride were introduced. The particle size of the plutonium oxide powders obtained was less than 15 μm. The powders showed poor flowability, but after granulation they were suitable for fabricating kernels with mixed fuel. The gallium was removed by reduction of Ga2O3 by hydrogen to Ga2O, which was sublimated. The mixed-fuel kernels sintered at 1600–1700°C in a hydrogen atmosphere contained <0.001 wt.% gallium, and their density was 94–97% of the theoretical value.

Neutron and gamma-ray cross-correlation measurements of plutonium oxide powder

For the first time, measurements of the time-dependent cross-correlation distributions of plutonium oxide have been made separately for neutrons and gamma rays. Six EJ-309 liquid scintillation detectors with a digital, offline pulse shape discrimination and pulse timing method were used to measure five different samples of varying mass and burnup. The number of (neutron, neutron) correlations were selectively analyzed versus plutonium mass and a clear, increasing trend was observed. Additionally, the measurement scenarios were modeled using the MCNP-PoliMi code and good agreement was observed between the measured and simulated cross-correlation functions.

Effective thermal conductivity of MOX raw powder

The radial temperature distribution of plutonium and uranium mixed oxide powder loaded into a cylindrical vessel was measured in air and argon gas, and the effective thermal conductivity was calculated from the measured temperature distribution and the decay heat. The effective thermal conductivities were small values of 0.061–0.13 W m –1 K –1 at about 318 K, and changed significantly with O/M, bulk density and atmospheric gas. The results in this work were analyzed by the model of Hamilton and Crosser and a new model for the effective thermal conductivity of the powder was derived as functions of powder properties and thermal conductivity of atmospheric gas.

Evaluation of safety of French type B package designs in severe accident environments other than regulatory

Abstract Packages used to transport radioactive materials in France must be designed to meet the safety performance requirements when subject to the test conditions set forth in the International Atomic Energy Agency (IAEA) Regulations. During actual use, the packages may be subject to quite different accident conditions. The Institut de Radioprotection et de Surete Nucleaire (IRSN) has evaluated the behaviour of typical packages designed to transport spent fuel, high activity waste, fresh mixed oxide (MOX) fuel and plutonium oxide powder under realistic conditions of mechanical impact and fire. The studied designs remain safe after impact onto targets present in the real environment of transport. The energy absorption by the package ancillary equipment (transport frame) compensates for the kinetic energy increase by comparison to the energy expended during the regulatory tests. New software was developed to correctly simulate the thermal behaviour of the neutron shielding materials. The studied package d...

Effects of pressure on thermal transport in plutonium oxide powder

Radial temperature profiles of plutonium dioxide powder in a cylindrical vessel were measured over a pressure range of 0.055 to 334.4 kPa with two different fill gases, helium and argon. The powder provides a very uniform self-heating medium for analysis. A thermal conductivity model was developed for heat conduction in the fine ceramic powder. Most literature models make limiting assumptions about powder characteristics that do not hold for this material. Despite the powder particles’ complex geometry, the proposed model correctly reproduces the powder temperature profiles over the wide pressure range for both fill gases.

Oxidation of Plutonium Metal

The processes leading to the production of plutonium oxide powders as a result of the interaction of plutonium metal with moist (5% H2O) air and moist (5% H2O) argon are investigated. The kinetic curves in the range 250–400°C are obtained. The oxidation is shown to occur in stages; the activation energy of each stage is calculated. A mechanism of the oxidation of plutonium is proposed.

The kinetics of plutonium oxide dissolution in nitric/hydrofluoric acid mixtures

The dissolution rate of plutonium oxide powder in nitric/hydrofluoric acid mixtures is controlled by a chemical reaction at the PuO2 surface rather than by diffusion. The reaction is first order with respect to both the surface area (S) of PuO2 and the concentration of free (non-complexed) hydrofluoric acid ([HF]ƒ) in solution. As the dissolution progresses, the surface area decreases and free HF is consumed by complexation with dissolved Pu(IV) to form PuFx(4−x)+ complexes. This results in a rapid decrease of the dissolution rate with time. The initial rate is not affected by HNO3 concentration in the range of 4–12 M. However, at longer reaction times the rate increases with increasing HNO3 concentration. At higher HNO3 concentrations more free HF is available for reaction since plutonium fluoride complexes are less stable. Rate equations which describe changes in S and [HF]ƒ during dissolution are derived and rate constants are given for dissolution of PuO2 obtained by burning the metal. The second order rate constant is 3.0 × 10−6l/sec-m2 at 35°C for dissolution of this oxide.