Uranium-plutonium Mixed Oxide Research Articles

Passive gamma-ray scanning for non-destructive characterisation of mixed oxide (MOX) fuel pins for BWRs

Uranium-plutonium mixed oxide (MOX) fuel is being fabricated at the Advanced Fuel Fabrication Facility, Bhabha Atomic Research Centre, Tarapur, as fuel for boiling water reactors (BWRs). The sintered fuel pellets are encapsulated in zircaloy-2 clad tubes using TIG welding of end plugs. Enrichment of plutonium dioxide is an important characteristic which needs to be assured by non-destructive means. The use of passive gamma-ray scanning for non-destructive monitoring of enrichment of plutonium dioxide in BWR MOX fuel rods is demonstrated. This paper describes the experiments on gamma-ray scanning of the welded MOX fuel rods carried out at AFFF, BARC and presents the results.

Analysis of Mixed Oxide Fuel Loaded Cores in the Heavy Water Reactor FUGEN

Uranium-plutonium mixed oxide (MOX) fuel cores in the heavy water reactor, FUGEN, were analyzed using the Advanced Thermal Reactor (ATR) type core design code system WIMS-ATR/POLESTAR and the accuracy of this code system also has been evaluated by means of operational data through the 34 burnup cycles and on-site γ-scanning data. The root mean square errors of calculated thermal neutron flux distributions were less than 5% compared with the power calibration monitor traverse data. The root mean square error of calculated power distributions was less than 4% compared with the γ-scanning data. The root mean square error of calculated burnup distributions was less than 3% compared with the γ-scanning data. The averaged effective multiplication factor was 1.000 and its standard deviation was 0.002. The calculation accuracy of void reactivity coefficient was ±3x 10-5 ∇k/k/%void for the equilibrium cores. The calculation accuracy of power coefficient was ±1.5x10-5∇k/ k/%power. The accuracy of ATR type core design code system was enough for the core management in the Fugen Nuclear Power Station.

Fabrication of Mixed Urianium–Plutonium Fuel on the Paket Setup and Waste Handling Methods

The results of fabrication of fuel elements with mixed uranium–plutonium oxide fuel are presented. The experimental fuel assemblies assembled from the fuel elements were tested in BN-350 and -600 reactors. Postreactor investigations of the fuel elements showed that there was no substantial difference in the behavior of the fuel cores consisting of the mixed fuel as compared with UO2 fuel. Solid and liquid radioactive wastes are produced during the fuel fabrication process. A classification of the wastes and methods for handling them is given. It is shown that the off-grade sintered pellets should be pulverized and returned to the beginning of the mixed-fuel fabrication process.

The irradiation test of Inert-Matrix Fuel in comparison to uranium plutonium mixed oxide fuel at the Halden Reactor

The irradiation test of Inert-Matrix Fuel in comparison to uranium plutonium mixed oxide fuel at the Halden Reactor

Design of an overmoderated fuel and a full MOX core for plutonium consumption in boiling water reactors

The use of uranium–plutonium mixed oxide fuel (MOX) in light water reactors (LWR) is nowadays a current practice in several countries. Generally 1/3 of the reactor core is loaded with MOX fuel assemblies and the other 2/3 with uranium assemblies. Nevertheless the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this paper the design of a boiling water reactor (BWR) core fully loaded with an overmoderated MOX fuel design is investigated. The design of overmoderated BWR MOX fuel assemblies based on a 10×10 lattice are developed, these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. In order to increase the moderator to fuel ratio two approaches are followed: in the first approach, 8 or 12 fuel rods are replaced by water rods in the 10×10 lattice; in the second approach, an 11×11 lattice with 24 water rods is designed with an active fuel length very close to the standard MOX assembly. The results of the depletion behavior and the main steady state core parameters are presented. The feasibility of a full core loaded with the 11×11 overmoderated MOX fuel assembly is verified. This design take advantage of the softer spectrum comparable to the 10×10 lattice with 12 water rods but with thermal limits comparable to the standard MOX fuel assembly.

The irradiation test of inert-matrix fuel in comparison to uranium plutonium mixed oxide fuel at the halden reactor

Conservative modelling for pin layout shows that the relatively low thermal conductivity of Inert-Matrix Fuel (IMF) causes higher temperatures and therefore higher fission gas release than in uranium plutonium mixed oxide (MOX). According to neutronic calculations, performance differences will also arise from different evolutions of the respective radial power and burnup distributions. Modelling of these effects as well as a ∼10% greater production of Xe in the thermal spectrum of the Halden reactor is well within the capabilities of appropriate codes. Some of the data and models used for the pre-calculations are preliminary and will be revised after the first experimental data have become available.

Development and Testing of Fuel Elements with Vibrationally Compacted Oxide Fuel for Fast Reactors

The main results of a series of scientific-research and technological studies performed at the State Science Center of the Russian Federation – Scientific-Research Institute of Nuclear Reactors to substantiate the use of fuel elements with vibrationally compacted oxide fuel in fast reactors are presented. In the course of this work, the physical-mechanical and technological characteristics of granular UO2 and UPuO2 fuel were studied; radiation tests and materials-engineering investigations of experimental and test fuel elements were performed in BOR-60, BN-350, and -600 reactors. More than 30,000 fuel elements were fabricated. Maximum burnup ∼30% heavy atoms was attained in BOR-60 using fuel assemblies with the standard construction and 32.3% heavy atoms was obtained using experimental fuel elements with a collapsible fuel assembly. In testing fuel elements with vibrationally compacted UPuO2 in BN-600, maximum burnup of 9.6% (∼10.8% heavy atoms for individual fuel elements) was achieved. Postreactor investigations showed that the service life of the fuel elements is determined only by the choice of the cladding material. In accordance with the concept developed at the Ministry of Atomic Energy of Russia for the utilization of weapons plutonium, the Institute set about to implement in practice a technology for converting the metallic weapons-grade plutonium into mixed uranium–plutonium oxide fuel on the basis of pyroelectrochemistry and vibrational compaction.

10.1007/BF02898464

During the fabrication of mixed uranium-plutonium oxide (MOX) fuel rods, two important characteristics to be checked in as-fabricated fuel pins are plutonium enrichment and plutonium dioxide agglomerates. The mixed oxide fuel pellets are made via mechanical mixing of uranium dioxide and plutonium dioxide powders by cold compaction and sintering. The chance of loading a wrong Pu enrichment pellet and having pellets with plutonium dioxide agglomerates in a fuel pin cannot be ruled out. A simple nondestructive evaluation technique is felt necessary to ensure at the last stage (in the welded pins) to check these two characteristics.

Determination of oxygen-to-metal atomic ratio in mixed oxide fuel by non-dispersive infrared spectrophotometry after fusion in an inert gas atmosphere

A new method of analysis for the determination of the oxygen-to-metal atomic ratio (O/M) in a uranium–plutonium mixed oxide (MOX) fuel was developed. The sample and the nickel metal were fused in a graphite crucible in a resistance heating furnace under helium flowing through the system. The oxygen in the MOX fuel reacted quantitatively with the crucible graphite producing carbon monoxide, which was determined using non-dispersive infrared spectrophotometry (NDIR). The oxygen content in the MOX fuel was calculated from the analysis results of the carbon monoxide. The O/M ratios determined by the present method were in good agreement with those of the thermogravimetric method. The relative standard deviation was less than 0.20% for sintered MOX fuel. Only 30 mg of sample was required, and the analysis time takes about 10 min. No specific skills for glove-box work are necessary. The present method is applicable to sintered MOX fuels over a wide range of plutonium contents.

Alternative versions of inert matrix fuel for the use of civil and weapons-grade plutonium in reactors

A fuel element approach that is an alternative to uranium–plutonium mixed oxide (MOX) fuel and inert matrix fuel is reviewed. It is designed to burn civil and weapons-grade plutonium in light water reactors (LWRs). The results of preliminary research reactor tests are given.

Behavior of metallic fission products in uranium–plutonium mixed oxide fuel

Metallic fission products, ruthenium, rhodium, technetium, palladium, and molybdenum, exist in irradiated oxide fuels as metallic inclusions. In this work, the radial distributions of metallic inclusion constituents in the fuel specimen irradiated to a peak burnup of 7–13 at.% were observed with an electron probe microanalysis. Palladium concentration is high at the periphery in all the specimens. Molybdenum shows the same tendency for the 13 at.% burnup specimen. These results showed the significant difference between experimental data and calculations with ORIGEN-2 at such high burnups, which suggested that the migration of palladium and molybdenum was controlled mainly by diffusion of gaseous species containing each metal along the fuel temperature gradient.

Distribution of molybdenum in FBR fuel irradiated to high burnup

Molybdenum is one of high yield fission products and has a chemical affinity for oxygen in uranium-plutonium mixed oxide fuel. In this work, radial distributions of molybdenum were investigated in high burnup fuels irradiated up to 13.33 at.%. It is found that the distributions are different from those expected from diffusion process in low burnup. This suggests that molybdenum in high burnup fuel migrates by not only a diffusion process but also by a gaseous molybdenum transport mechanism.

Determination of nitrogen in uranium-plutonium mixed oxide fuel by gas chromatography after fusion in an inert gas atmosphere

A gas chromatographic technique has been developed for the determination of nitrogen in uranium-plutonium mixed oxide fuel after fusion in an inert gas atmosphere. When the sample and pure iron powder in a graphite crucible were heated to approximately 2500 °C by a resistance heating furnace, a large amount of carbon monoxide was evolved with a small amount of nitrogen and hydrogen. A gas chromatograph equipped with a pre-cut system was used for the separation of nitrogen from the carbon monoxide. Nitrogen separated by the gas chromatograph was determined by means of a thermal conductivity detector. Only 100 mg of the sample was used, and the analysis requires about 10 min. No specific skills for glove-box work are necessary. The relative standard deviation and detection limit (3σ-criterion) were less than 5% and 9μg g −1, respectively. The present method is not only applicable to the analysis of research samples but also to the quality control of uraniumplutonium mixed oxide fuel production lines.

Design of highly moderated pressurized water reactor based on critical heat flux considerations

The difficulty of applying the existing critical heat flux correlations to the design of a modified 100% mixed plutonium uranium oxide fuelled assembly are outlined. A core with increased moderating ratio (RMA) was designed. The purpose of this design modification is to permit the consumption of large amounts of plutonium which cannot be burned in a conventional pressurized water reactor (PWR) owing to safety considerations. The design criteria required that the minimum departure from nucleate boiling ratio (DNBR) in the modified fuel does not exceed the value of the standard N4 PWR, which served the basis for the design. The desired conditions were achieved by reducing the fuel pin diameter to increase the moderating ratio and increasing the number of grids with mixing vanes to improve the minimum DNBR in the modified assembly. The design methodology and some of the proposed design options are presented.

Characterization of spent nuclear fuels by ion chromatography–inductively coupled plasma mass spectrometry

Ion chromatography (IC) coupled to ICP-MS was applied to the characterization of two different spent fuel samples [uranium oxide (UO2) and mixed uranium–plutonium oxide (MOX)] and the results were compared with those obtained by other techniques. Isotope dilution analysis with ion chromatographic separation was applied to the determination of the fission products (Rb, Sr, Cs, Ce, Nd, Sm, Eu and Gd). The standard additions method was employed for the determination of monoisotopic fission products and actinides (Y, La, Pr, 147Pm, 237Np, 241Am, 243Am and 244Cm). Total U and Pu were determined only by ID-TIMS. Nd, Am and Cm isotope concentrations were determined also by ID-TIMS. 144Ce, 154Eu, 134Cs and 137Cs were determined in parallel by gamma spectrometry. Based on the ID-TIMS results for U, Pu and 148Nd, fuel burn-up was calculated and the value introduced into the computer code KORIGEN in order to calculate the complete fuel inventory based on the known irradiation parameters for the two fuels. The agreement between the experimental IC–ICP-MS results and the theoretical calculations was within 15% for most isotopes.

Determination of Americium in a Nitric Acid Medium Using Second-Order Derivative Spectrophotometry.

A second-order derivative spectrophotometric method has been developed for the determination of microgram quantities of trivalent americium in a nitric acid medium containing plutonium and uranium. The second-order derivative absorbance of the americium at 503.5nm is used for the quantitative analysis. In the determination of 5.8μg/ml and 3.9μg/ml of americium, the presence of a 100-fold amount of plutonium and a 600-fold amount of uranium could be tolerated, respectively. The trivalent americium, in the concentration range of 1-11μg/ml has been determined with good precision. The detection limit was 0.3μg/ml at 3 times the standard deviation (3σ). The relative standard deviation was less than 2.1% for a sample of uranium-plutonium mixed oxide. The proposed method does not require the complicated separation step of plutonium and uranium.

Determination of the Modified Conversion Ratio of Light-Water-Moderated Uranium-Plutonium Mixed-Oxide-Fuel Lattice

The modified conversion ratio (MCR) (the ratio of the {sup 238}U capture rate to the total fission rate) in a light-water-moderated uranium-plutonium mixed-oxide (MOX-) fuel lattice was measured for four types of lattices with different plutonium enrichment. In the current method, the relative reaction rates of {sup 238}U capture and total fission were obtained from nondestructive gamma-ray spectrometry of {sup 239}Np and fission products, respectively, which accumulated in the fuel rod irradiated at the Tank-Type Critical Assembly. The measured results of the fission rates derived from two different fission products agreed well with each other, and the measured MCRs showed good agreement with the results of the Monte Carlo calculation with the whole-core model. Therefore, the current nondestructive method is applicable to the MCR measurement of MOX fuel.

Determination of occluded gases in mixed oxide fuel pellets using gas sampling unit and gas chromatography

A new technique for the determination of occluded gases in sintered plutonium-uranium mixed oxide (MOX) fuel pellets was developed. The gases extracted from the sample at vacuum condition were injected into a gas chromatograph through an originally developed gas sampling unit. The gases were simultaneously determined using the improved gas chromatograph equipped with several separation columns and a photoionization detector. The relative standard deviations ( n = 10) were 1.0, 3.4, 0.6, 1.0, 1.4, 0.8 and 0.9% for H 2, O 2, N 2, CH 4, CO 2, CO and C 2H 6, respectively. The gases H 2, O 2, N 2, CH 4, CO 2 and CO from the MOX fuel pellets were simultaneously determined in this study. The results of gas analysis using this method were in agreement with that of high temperature vacuum extraction method in the range less than 300 μl of extracted gas.

95/01071 The economics of the nuclear fuel cycle

The current international perception of future nuclear fuel cycle costs is reviewed and it is concluded that they should remain stable or decrease slightly over the next decade barring major discontinuities in the supply or acceptability of fossil fuels. The economic position of mixed plutonium-uranium oxide fuel is considered in relation to plutonium availability and price

Radial plutonium and fission product isotope profiles in mixed oxide fuel pins evaluated by secondary ion mass spectrometry

In the frame of the PRIMO programme the irradiation behaviour of uranium-plutonium mixed oxide fuel in pressurized water reactors is investigated. Secondary ion mass spectrometry was applied for the determination of radial profiles of burnup and of plutonium and fission product isotopes, providing data for the analysis of mixed oxide neutronic behaviour.