MOX Pellets Research Articles

Sintering and microstructural behaviors of mechanically blended Nd/Sm-doped MOX

ABSTRACT Additive MOX pellets are fabricated by mechanically blending the additive material powder and MOX powder. Nd2O3 and Sm2O3 are chosen as the additive materials to simulate the corresponding soluble fission products dispersed in MOX. Shrinkage curves of the MOX pellets are obtained by dilatometry, which reveal that the sintering temperature is shifted toward a value higher than that of the respective regular MOX. The additives, however, promote grain growth and densification, which can be explained by the effect of oxidized uranium cations to a pentavalent state. Ceramography reveals large agglomerates after sintering, and Electron Probe Micro-Analysis confirms that inhomogeneous elemental distribution, whereas XRD reveals a single face-centered cubic phase except for Sm-doped specimens indicating a small peak of Sm2O3. Finally, by grinding and re-sintering the Nd2O3-doped MOX, the cation distribution homogeneity is significantly improved, which can simulate spent nuclear fuels with soluble fission products.

Dissolution behavior of fast reactor MOX nuclear fuel pellets in nitric acid medium

The fundamental information required for designing a continuous FBR MOX spent fuel dissolution system is the intrinsic kinetics of dissolution of the MOX fuel under typical PUREX process conditions. The influence of parameters such as Pu content of the MOX pellet, nitric acid concentration, intensity of mixing and reaction temperature, on the dissolution behaviour of PFBR MOX fuel pellets in nitric acid was investigated. The results clearly bring out the preferential rapid dissolution of UO 2 over PuO 2 from the same MOX pellet. This behavior was found to be inversely proportional to all the above-mentioned influencing parameters except for the Pu content of the pellets in which case it is directly related. These results will pave way for better understanding of the dissolution mechanism of MOX fuel under typical PUREX process conditions.

Fabrication of homogenous pellets by freeze granulation of optimized TiO2-Y2O3 suspensions

MOX fuels are used in nuclear reactors. To ensure the manufacture of MOX, wet processes are investigated. Freeze granulation of water-based powder suspension seems a promising way to yield homogenous and easy-to-press UO2-PuO2 granules. These granules are expected to form dense and homogenous pellets by uniaxial pressing. Granules and pellet properties are affected by suspension formulation and atomization conditions. Therefore, these conditions must be studied and optimized to produce granules with good processability and thus MOX pellets with compliant density, homogeneity and absence of defects. In this scope, key properties of water-based suspensions of UO2 and PuO2 surrogate powders, TiO2 and Y2O3 respectively, were firstly studied to assess their suitability for the freeze granulation process. These properties were compared to those of UO2 and PuO2 suspensions to verify and confirm the correctness in the choice of surrogate powders. Then, the freeze granulation process itself was investigated focusing on TiO2-Y2O3 suspensions.

Validation of Burnup Calculation Code SWAT4 by Evaluation of Isotopic Composition Data of Mixed Oxide Fuel Irradiated in Pressurized Water Reactor

Abstract This paper provides validation results of SWAT4, the revised version of SWAT3.1, by the analysis of PIE data of MOX fuel irradiated in a pressurized water reactor. SWAT4 is a system that combines the point burnup system ORIGEN2 and neutron transport calculation solvers, the continuous energy Monte Carlo code MVP or MCNP5, and the deterministic neutronics calculation code SRAC2006. The calculation results of SWAT4 have generally same trends with the case of UO2 fuel analyses. For major uranium and plutonium isotopes, deviations less than 5% were obtained. This means that SWAT4 has the same accuracy to predict isotopic compositions of irradiated MOX fuel with the case of UO2 fuel. The radial distribution of isotopes in a pellet was also analyzed, whose results were compared with that measured by SIMS. SWAT4 predicted well the isotope and burnup distributions in an irradiated MOX pellet.

The Misuse of Spreadsheets in the Nuclear Fuel Industry

This paper considers the management, technological and human factor issues that led to the BNFL fuel rod spreadsheet data falsification incident in 1999. BNFL discovered in 1999 that some data supporting quality assurance and safety processes had been falsified by BNFL workers using spreadsheets. The implication of this finding was that some of the Mixed Oxide Fuel Pellets shipped to customers in Japan for use in nuclear reactors were of an unknown mass and quality. This meant that the use of the MOX pellets fuel produced by BNFL would introduce uncontrolled factors into the safe operation of nuclear reactors. This paper will examine the production of MOX pellets at the Sellafield site, the falsification of data and the report commissioned by HM Nuclear Inspectorate. The paper will then identify a number of managerial and technological failings that led BNFL to use spreadsheets for recording such data. Finally the paper analyses other cases of spreadsheet fraud and explores some strategies for reducing the likelihood and impact of spreadsheet errors and fraud.

Strength loss in MA-MOX green pellets from radiation damage to binders

The fracture strength of green Minor Actinides (MA)-MOX pellets containing 75wt.% DUO2, 20wt.% PuO2, 3wt.% AmO2 and 2wt.% NpO2 was studied as a function of storage time, after mixing with the binder and before sintering, to test the effect of radiation damage on binders. Fracture strength degraded continuously over the 10days of the study for all three binders studied: PEG binder (Carbowax 8000), microcrystalline wax (Mobilcer X) and styrene–acrylic copolymer (Duramax B1022) but the fracture strength of Duramax B1022 degraded the least. For instance, for several hours after mixing Carbowax 8000 with MA-MOX, the fracture strength of a pellet was reasonably high and pellets were easily handled without breaking but the pellets were too weak to handle after 10days. Strength measured using diametral compression test showed that strength degradation was more rapid in pellets containing 1.0wt.% Carbowax PEG 8000 compared to those containing only 0.2wt.%, suggesting that irradiation not only left the binder less effective but also reduced the pellet strength. In contrast the strength of pellets containing Duramax B1022 degraded very little over the 10days period. It was suggested that the styrene portion present in the Duramax B1022 copolymer provided the radiation resistance.

The Chemistry of (U,Pu)O2 Dissolution in Nitric Acid

The focus of this paper is the chemistry of mixed uranium plutonium oxide (MOx,) in nitric acid. An overview of dissolution chemistry is discussed by comparing the differences in the dissolution characteristics of uranium and plutonium oxides. An overview of batch dissolution experiments, studying the dissolution chemistry of high surface area MOx powders and low surface area MOx pellets with reference to the effects of nitrous acid, nitric acid and temperature are described. The results are discussed in terms of the autocatalytic mechanism and mass transfer limited dissolution.

Flowability measurement of pulverized and granulated materials using vibrating tube method

To develop a method for measuring the flowability of MOX (mixed oxide of uranium and plutonium) particles used in the simplified MOX pellet fabrication process, the flowability of model particles has been investigated by the vibrating tube method. As model particles, pulverized ZrO 2 and granulated WO 3 were used. To prepare a variety of samples, coarse particles of 106–250 μm in diameter were mixed with fine particles smaller than 45 μm in diameter in different concentrations. The prepared particles were put into the vibrating tube and the amplitude of vibration was increased at a constant rate for a period of time and then decreased. The mass of particles discharged from the tube was measured at constant time intervals. From the experimental data, the relationships between the mass flow rate and the vibration acceleration (i.e. flowability profiles) were obtained. Two factors (i.e. ‘critical vibration acceleration’ to make the particles flow and ‘characteristic mass flow rate’) were selected to analyze the flowability profiles. The hysteresis of the flowability between increasing and decreasing vibration accelerations was also evaluated.

Characteristics and Sinterability of MOX Powder Prepared by the Microwave Heating Denitration Method

The characteristics and sinterability of UO2-PuO2 mixed oxide (MH-MOX) powder prepared by the microwave heating denitration method were measured and compared with those of UO2 (ADU-UO2) powder prepared by the ADU method. Furthermore, the degree of surface roughness and flowability of MH-MOX powder were evaluated and also compared with those of ADU-UO2 powder. The degree of surface roughness of ADU-UO2 powder calcined at temperatures >700°C significantly decreased, and its sintered density also dropped below 80% theoretical density. However, the degree of surface roughness and sinterability of MH-MOX powder calcined at 950°C were higher than those of ADU-UO2 powder. These results could be understood using the concept of Hüttig and Tamman temperatures, which is commonly cited for ceramic materials. The flowabilities of MH-MOX and ADU-UO2 powders decreased with an increase of compressibility, and they were categorized as non-free-flowing according to Carr’s theory on powder flowability. It is, therefore, necessary for the mixed powder of MH-MOX powder, ADU-UO2 powder, and dry recycled MOX scrap powder to be granulated to provide a free-flowing feed to the pelletizing press in the MOX pellet fabrication process.

Fuel, Core Design and Subchannel Analysis of a Superfast Reactor

A compact supercritical water-cooled fast reactor (superfast reactor) core with a power of 700MWe is designed by using a three-dimensional neutronics thermal-hydraulic coupled method. The core consists of 126 seed assemblies and 73 blanket assemblies. In the seed assemblies, 251 fuel rods, consisting of MOX pellets, stainless steel (SUS304) cladding, and fission gas plenum are arranged into a tight triangle lattice along with 19 guide tubes for control rods and instrumentation. A zirconium hydride (ZrH) layer is employed in the blanket assemblies to reduce void reactivity. The results of the coupling three-dimensional neutronics and thermal hydraulic calculations show that this core has a high power density of 158.8 W/cm3 with a maximum linear heat generation rate (MLHGR) less than 39 kW/m, that an average coolant outlet temperature of 500°C is achieved with a maximum cladding surface temperature (MCST) less than 650°C, and that void reactivity coefficients are negative throughout the cycle. Since the thermal-hydraulic part of the core design is based on single-channel analyses, subchannel analyses are also performed on all the seed assemblies to clarify the influence of cross-flow.

Fuel, Core Design and Subchannel Analysis of a Superfast Reactor

A compact supercritical water-cooled fast reactor (superfast reactor) core with a power of 700MWe is designed by using a three-dimensional neutronics thermal-hydraulic coupled method. The core consists of 126 seed assemblies and 73 blanket assemblies. In the seed assemblies, 251 fuel rods, consisting of MOX pellets, stainless steel (SUS304) cladding, and fission gas plenum are arranged into a tight triangle lattice along with 19 guide tubes for control rods and instrumentation. A zirconium hydride (ZrH) layer is employed in the blanket assemblies to reduce void reactivity. The results of the coupling three-dimensional neutronics and thermal hydraulic calculations show that this core has a high power density of 158.8 W/cm3 with a maximum linear heat generation rate (MLHGR) less than 39 kW/m, that an average coolant outlet temperature of 500°C is achieved with a maximum cladding surface temperature (MCST) less than 650°C, and that void reactivity coefficients are negative throughout the cycle. Since the the...

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).

Developments in the fabrication technology of low density MOX pellets for fast breeder reactor fuel

The fabrication technology of low density MOX pellets was studied using some organic compounds as pore formers. The pore former, K3 showed the best performance among five organic compounds in the 7.5 kg-UOX runs, but K3 lost this in the 36 kg-MOX run, possibly because plutonium decay heat affected its performance. Comparison of thermal stability between K3 and a newly introduced pore former, Avicel, of high softening temperature, showed that K3 lost its spherical particle shape and consequently, its pore forming ability at 70 °C which was below its reported melting point (84–88 °C) while Avicel could maintain its spherical particle shape up to 150 °C. The pore forming performance of Avicel was also confirmed in the 36 kg-MOX run which is the same scale as for mass production of MOX fuel pellets.

Effect of residual carbon on the sintering behavior of MOX pellets

Sintering behavior of MOX compacts containing different amounts of carbon (290–1735 ppm) was investigated at temperatures from 1073 to 2023 K to study the effect of residual carbon. Specimen shrinkages were measured with a non-contact type optical dilatometer during heating. The compact shrinkage at temperatures from 1273 to 1523 K was about six times larger in the two specimens with high residual carbon content than in the two with low content. This behavior could be understood by considering that shrinkage of the former specimens was enhanced in this temperature range by the decrease of O/M ratio due to the evolution of CO gas and consequently the significant increase of the metal ion inter-diffusion coefficient. In the specimens with low residual carbon content, the amount of CO gas evolution was too small to affect the inter-diffusion coefficient. This difference in shrinkage between the two kinds of specimens was also discussed from a theoretical model applied to the initial sintering stage of ceramics.

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

To assess the feasibility of the 31% Pu-MOX fuel rod design of reduced-moderation water reactor (RMWR) in terms of thermal and mechanical behaviors, a single rod assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO 2 fuel have been used as appropriate. The results are: fission gas release is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400 K, while cladding diameter increase caused by pellet swelling is within 1% strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling and cladding oxidation require to be investigated in detail.

Lithium reduction of a MOX pellet

The MOX pellet of approximately 10g was reduced with lithium metal in molten lithium chloride at 923K. The pellet has high density and consists of 90% of uranium, 10% of plutonium, and 0.2% of americium growing in from plutonium. No major fracturing of the pellet occurred during reaction. Uranium and plutonium were essentially completely reduced to metal with no significant change in the ratio between them. Most of the lithium oxide generated by the reaction dissolved in the molten salt. SEM observation clearly showed a lot of pores and cracks in the reacted pellet. The generation of the apertures was caused by significant decrease of molar volume from oxide to metal. Plutonium and americium were detected in the samples taken from the molten salt during the experiment. Approximately 5% of initial load of plutonium and 75% of americium eluded from the pellet and most of them precipitated on the bottom of the crucible in end of the experiment.

Failure behavior of plutonium-uranium mixed oxide fuel under reactivity-initiated accident condition

Two series of in-pile tests on MOX fuels were performed in the NSRR to study failure behavior under RIA (reactivity-initiated accident) conditions in water cooled reactors. PWR type MOX test rods were pulsed in a first series. The test rods were designed to have dimensions identical to standard UO 2 fuel, on which a large number of tests had been conducted previously. The test result was that the failure mechanism and the threshold of MOX fuel was consistent with those of UO 2 fuel. ATR-type MOX test rods with PuO 2 particles as well as reference rods without PuO 2 particles were subjected to pulsing in a second series. PuO 2 particles of 400 and 1100 μm in diameter were artificially embedded at the surface of MOX pellets. No effect of particles appeared on the threshold, and no significant indication of their effect was observed on the cladding.

Oxidation and pulverization of UO 2 and mox pellets

Accidents in plutonium processing plants due to the combustion of kerosene may involve the oxidation and pulverization of nuclear fuel pellets. Hence, some simulation experiments were performed in a glove box to study the behaviour of pellets of different composition ( UO 2 PuO 2 , i.e. MOX) and preparation conditions (sintered or green pellets with or without binding agent) as a function of temperature, time and oxidation atmosphere, especially in relation to the formation of coarse and respirable particles.

Bestimmung der Pu-Löslichkeit an unbestrahlten Pu, U-Mischoxid-Kernbrennstoffen

A quantitative analytical method for the determination of the solubility of unirradiated (Pu, U)O 2 fuels is described below. The results indicate, that the solubility values are mainly depending on the processing variables, the densitiy of the MOX pellets, the time of dissolution and the concentration of the nitric acid. The definition of Pu-solubility for the standardised fuel specification and a suitable analytic procedure are still open to further discussions.