MOX Fuel Research Articles

Fuel clad chemical interactions in fast reactor MOX fuels

Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements (oxygen, cesium, tellurium, iodine) in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction (FCCI) models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL (Hanford Engineering Development Laboratory) relation is recommended: d/μm=({0.507⋅[B/(at.% fission)]⋅(T/K-705)⋅[(O/M)i-1.935]}+20.5) for (O/M)i⩽1.98. A new model is proposed for (O/M)i⩾1.98: d/μm=[B/(at.% fission)]⋅(T/K-800)0.5⋅[(O/M)i-1.94]⋅[P/(Wcm−1)]0.5. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, (O/M)i is the initial oxygen-to-(uranium+plutonium) ratio, and P is the linear power rating. For fuels with [n(Pu)/n(M=U+Pu)]>0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.

Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths.

Thermal conductivity of heterogeneous LWR MOX fuels

It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3–15 and 15–30wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation.

Thermal diffusivity measurement of (U, Pu)O2−x at high temperatures up to 2190 K

The thermal diffusivities of uranium and plutonium mixed oxide (MOX) fuels with 30% Pu-content were measured at temperatures from 990 to 2190K by the laser flash method. In high-temperature measurements of the stoichiometric specimen, a difference was observed between measurements during heating up and cooling down which seemed to be caused by a change of the oxygen to metal (O/M) ratio. Estimation of the O/M ratio decline by the Knudsen–Langmuir equation, which can be used to discuss evaporation behavior, suggested that the O/M ratio of the stoichiometric specimen had started to decrease at 1800K. For the O/M ratio of 1.95, that estimation showed that the O/M ratio did not change up to 2200K. It was also found from the experimental results that no significant changes of O/M ratios of specimens with O/M ratio less than 1.95 were observed.

New Fresh Mox Fuel Packages: From Design to Operation

AbstractThe MX packages developed by COGEMA Logistics according to TS-R-1 requirements will replace older Current packaging to transport fresh pressurised water reactor (PWR) and boiling water reactor (BWR) mixed oxide (MOX) fuels in Europe. Two types of package have been developed: (1) MX8 with a capacity of eight 17 × 17 900 MWe PWR fuel assemblies for dry loading and underwater unloading operations; and (2) MX6 for dry loading and unloading operations. The capacity of the MX6 is six 16 × 16 or 18 × 18 PWR fuel assemblies or sixteen 10 × 10 BWR fuel assemblies. To meet these capacities requirements, an innovative and optimised design has led to ‘mid-weight’ packages with original solutions for the body, the baskets and the fuel restraining system. To cope with both capacity and legal weight transport requirements, a new high-security transport system has been developed simultaneously. The first shipment with MX8 was made in December 2001, and the first use of MX6 packages is scheduled for the end of 2003.

ハルデン炉での高燃焼度MOX燃料照射試験:(その3) FPガス放出挙動に及ぼすMOX燃料製造法の影響

ハルデン炉での高燃焼度MOX燃料照射試験:(その3) FPガス放出挙動に及ぼすMOX燃料製造法の影響

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.

Behavior of Transuranic Mixed-Oxide Fuel in a CANDU-900 Reactor

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation, and heat loads of spent material. The burning of transuranic (TRU) fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while significantly reducing the fast reactor infrastructure needed. This paper examines the features of actinide mixed-oxide (MOX) fuel, TRUMOX, in a CANDU® nuclear reactor. The actinide concentrations used were based on extraction from 30-year-cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, supercell calculations were analyzed in DRAGON, and full-core analysis was executed in the RFSP two-group diffusion code. A time-average full-core model was produced and analyzed for reactor coefficients, reactivity device worth, and online fueling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 29.91 MWd/kg heavy element and an actinide transmutation rate of 35% for a single pass. A fully TRUMOX-fueled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing, and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle.

Curium analysis in plutonium uranium mixed oxide by x-ray fluorescence and absorption fine structure spectroscopy

Plutonium uranium mixed oxide (MOX) fuels are being used in commercial nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regards to their environment and the coolant. In this work the study of the local occurrence, speciation and next-neighbour environment of curium (Cm) in the (Pu,U)O2 lattice within an irradiated (60MWdkg−1 average burn-up) MOX sample was performed employing micro-x-ray fluorescence (µ-XRF) and micro-x-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Cm (∼0.7wt% in the rim and ∼0.03wt% in the centre) are determined from the experimental data gained for the irradiated fuel material examined in its centre and peripheral (rim) zones of the fuel. Curium occurrence is also reduced from the centre (hot) to the periphery (colder) because of the condensation of these volatile oxides. In the irradiated sample Cm builds up as Cm3+ species (>90%) within a [CmO8]13− or [CmO7]11− coordination environment and no (<10%) Cm(IV) can be detected in the rim zone. Curium dioxide is reduced because of the redox buffering activity of the uranium dioxide matrix and of its thermodynamic instability.

Coupled neutronics/thermal hydraulics evaluation for thorium based fuels in thermal spectrum SCWR

Thorium can supplement the current limited reserves of uranium. In current study, analyses are performed for thorium based fuels in thermal neutron spectrum Super Critical Water Reactor (SCWR). Thorium based fuels are studied in two roles. First role being replacement of conventional uranium dioxide fuel while the other being burner of Reactor Grade Plutonium (RG-Pu) in thermal neutron spectrum SCWR. Coupled neutron physics/thermal hydraulics analyses are performed due to large density variation of coolant over the active fuel length. Analyses reveal that thorium-uranium MOX fuels lead to smaller burnup values as compared to equivalent enriched uranium dioxide but possess the advantage of smaller excess reactivity at Beginning of Life (BOL). This can lead to savings in the form of Burnable Poisons (BP). Smaller fuel average temperature values are obtained for thorium-uranium MOX fuels as compared to uranium dioxide fuel option. Coated fuel option utilizing mixed thorium-uranium mono nitride fuel can help further decrease fuel average temperature values for thorium based fuels. U-233, produced in thorium uranium fuels, contribution towards fission energy produced is smaller as compared to plutonium produced in conventional uranium dioxide fuel. In terms of proliferation resistance, approximately 40% less quantity of plutonium is produced for thorium-uranium MOX fuels (for studied compositions) as compared to equivalent enriched uranium dioxide fuel. But, there is not much difference between the discharged plutonium vector compositions. Thorium–Plutonium based fuels lead to significantly harder spectrum which results in larger spread in radial power density and eventually causes larger values for thermal hydraulic parameters like fuel and clad temperature. Due to almost no production of plutonium, thorium based fuels can be a very good option to burn RG-Pu in thermal spectrum SCWR. Thorium based fuels destroyed almost 74% initially loaded RG-Pu as compared to 60% for uranium based MOX. HEU based thorium fuels can be a very good option for replacing conventional uranium dioxide fuels as very small quantities of plutonium is produced. This option, although, has regulatory issues due to use of HEU material.

Synthesis of the VERCORS experimental programme: Separate-effect experiments on Fission Product release, in support of the PHEBUS-FP programme

The VERCORS programme is an experimental programme composed of separate-effect tests specifically focusing on determining the Source Term (i.e. quantifying the nature and the release rate of fission products (FP) and actinides) from an irradiated fuel rod sample, under conditions representative of those encountered during a severe Light Water Reactor (LWR) accident. These tests, financed jointly by the French Institute for Radiological Protection and Nuclear Safety (IRSN) and the French electric utility EDF, were carried out in a specific high-activity cell at the CEA Grenoble Centre (French Alternative Energies and Atomic Energy Commission), and were commonly considered to be complementary to the PHEBUS-FP integral programme. They are comparable with certain tests carried out abroad, such as HI/VI in the USA, VEGA in Japan and the programme conducted at the CRL in Canada. Following on from the HEVA programme, VERCORS was conducted in a similar manner over a 14-year period between 1989 and 2002, using an experimental set-up which continued to evolve over time. A total of 17 tests were performed in accordance with three main phases. A first series of six tests (VERCORS-1 to -6) was conducted between 1989 and 1994 on UO2 fuel samples up to 2600K. Next, two different test series – VERCORS HT (three tests) and RT (eight tests) – were alternately performed between 1996 and 2002 at higher temperature, up to the fuel sample melting. These tests focused on UO2 and MOX fuels with a variety of initial configurations (intact or debris beds). The instrumentation become progressively more sophisticated, thus making it possible to extend the knowledge database to include the transport of FP (HT loop), or on the contrary focus on the release objective by associating the study of actinides (RT loop).With the exception of two tests (in debris bed configuration), the fuel sample was composed of a Pressurised Water Reactor (PWR) section containing three pellets in their original cladding. This sample was obtained from a father rod irradiated for several years in one of EDF’s nuclear power plants. In most of the tests, this sample was re-irradiated for several days in a Material Testing Reactor (MTR) of the CEA (SILOE or OSIRIS) so as to rebuild the inventory of FPs with short half-lives.This programme made it possible to quantify precisely fission product releases in all the situations explored, as well as to identify similar behavioural patterns amongst some of these fission products, thus making it possible to classify them schematically into four groups with decreasing volatility:•Volatile FP including fission gases, iodine, caesium, antimony, tellurium, cadmium, rubidium and silver with very high releases (practically total release) at temperatures of around 2600K.•Semi-volatile FP, a category composed of molybdenum, rhodium, barium, palladium and technetium with releases of 50–100%, but very sensitive to oxygen potential and with marked deposits in the vicinity of their emission point.•FP that are low volatile, such as ruthenium, cerium, strontium, yttrium, europium, niobium and lanthanum, with significant releases of around 3–10% on average, but capable (for some elements and under particular conditions) of reaching 20–40%.•Non-volatile FP composed of zirconium, neodymium and praseodymium, for which no release can be measured by gamma spectrometry for the envelope conditions of the VERCORS test grids.Actinides each have their own type of behaviour. They can nevertheless be subdivided into two categories, the first including U and Np, with releases of up to 10% and behaviour similar to that of the low volatile FP, and the second (Pu) with very low releases, typically less than 1%.

Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O₂ lattice in an irradiated (60 MW d kg⁻¹) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (~0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am³⁺ species within an [AmO₈]¹³⁻ coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix. - Graphical abstract: Americium LIII XAFS spectra recorded for the irradiated MOX sub-sample in the rim zone for a 300 μm×300 μm beam size area investigated over six scans of 4 h. The records remain constant during multi-scan. The analysis of the XAFS signal shows that Ammore » is found as trivalent in the UO₂ matrix. This analytical work shall open the door of very challenging analysis (speciation of fission product and actinides) in irradiated nuclear fuels. - Highlights: • Americium was characterized by microX-ray absorption spectroscopy in irradiated MOX fuel. • The americium redox state as determined from XAS data of irradiated fuel material was Am(III). • In the sample, the Am³⁺ face an AmO₈¹³⁻coordination environment in the (Pu,U)O₂ matrix. • The americium dioxide is reduced by the uranium dioxide matrix.« less

Molecular Dynamics study of the mixed oxide fuel thermal conductivity

There is still no clear understanding of the plutonium content influence on the thermal conductivity behaviour of the (U,Pu) O2 MOX fuels. In this work Classical Molecular Dynamics (MD) was used to investigate the (U,Pu) O2 thermal conductivity in the whole concentration range and in the temperature range from 400K to 1600K. The Green–Kubo approach was used for the thermal conductivity calculation and an algorithm was proposed to improve the accuracy of the calculation. The obtained results are in good agreement with the literature experimental data and results of modelling of other authors. On the basis of the obtained results we give recommendations for the MOX thermal conductivity evaluation in the concentration range from pure UO2 up to pure PuO2.

LWR spent fuel transmutation with fusion-fission hybrid reactors

In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.

Americium and plutonium release behavior from irradiated mixed oxide fuel during heating

The release behavior of Pu and Am was investigated under the reducing atmosphere expected in sodium cooled fast reactor severe accidents. Irradiated Pu and U mixed oxide fuels were heated at maximum temperatures of 2773K and 3273K. EPMA, γ-ray spectrometry and α-ray spectrometry for released and residual materials revealed that Pu and Am can be released more easily than U under the reducing atmosphere. The respective release rate coefficients for Pu and Am were obtained as 3.11×10−4min−1 and 1.60×10−4min−1 at 2773K under the reducing atmosphere with oxygen partial pressure less than 0.02Pa. Results of thermochemical calculations indicated that the main released chemical forms would likely be PuO for Pu and Am for Am under quite low oxygen partial pressure.

Restructuring and redistribution of actinides in Am-MOX fuel during the first 24h of irradiation

Abstract In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95.

Dissolution behavior of irradiated mixed oxide fuel with short stroke shearing for fast reactor reprocessing

An efficient dissolution process was established for future reprocessing in which mixed-oxide (MOX) fuels with high plutonium contents and dissolver solution with high heavy-metal (HM) concentrations (more than 500 g dm−3) will be treated. This dissolution process involves short stroke shearing of fuels (∼10 mm in length). The dissolution kinetics of irradiated MOX fuels and the effects of the Pu content, HM concentration, and fuel form on the dissolution rate were investigated. Irradiated fuel was found to dissolve as 102–103 times fast as non-irradiated fuel, but the rate decreased with increasing Pu content. Kinetic analysis based on the fragmentation model, which considers the penetration and diffusion of nitric acid through fuel matrices prior to chemical reaction, indicated that the dissolution rate of irradiated fuel was affected not only by the volume ratio of liquid to solid (L/S ratio) but also by the exposed surface area per unit mole of nitric acid (A/m ratio). The penetration rate of nitric acid is expected to be decreased at high HM concentrations by a reduction in the L/S ratio, but enhanced by shearing the fuel pieces with short strokes and thus enlarging the A/m ratio.

Backscatter gauge description for inspection of neutron absorber content

This paper describes design, calibration, and testing of a dual He-3 detector neutron backscatter gauge for use in the Savannah River Site Mixed Oxide Fuel project. The gauge is demonstrated to measure boron content and uniformity in concrete slabs used in the facility construction.

Study of Plutonium Recycling Scheme with Uranium Multi-recycling System

We have considered a fuel cycle scheme for multi-recycling of plutonium recovered from spent UO2 fuels of LWRs. Plutonium is to be partially loaded into the next LWRs in the form of MOX fuels with UO2 fuels made by re- enrichment of uranium recovered from spent fuels of LWRs. We consider to recycle plutonium in combination with uranium multi-recycling system previously reported, which can provide uranium fuels for not only fully UO2 fueled PWRs but also partially MOX fueled PWRs. The results of the process system analysis suggest that the proposed scheme makes it possible to recycle plutonium in succession.

Burn-Up Measurements on Dissolver Solution of Mixed Oxide Fuel Using HPLC-Mass Spectrometric Method

Burn-up measurement on an irradiated mixed oxide (MOX) test fuel pellet was carried out through measurements on the dissolver solution by HPLC-Thermal Ionization Mass Spectrometric (TIMS) technique. The studies carried out using HPLC as well as TIMS for quantification of burn-up value are described. While in one case, both the separation and determination of elements of interest (U, Pu and Nd) were carried out by HPLC; in another case, TIMS technique was used to quantify them from the HPLC separated fractions.The rapid separation procedures developed in our laboratory earlier were employed to isolate pure fractions of the desired elements. The individual lanthanide fission products (La to Eu) were separated from each other using dynamic ion-exchange chromatographic technique whereas uranium and plutonium were separated from each other using reversed phase chromatographic technique. The pure fractions of U, Pu and Nd obtained after HPLC separation procedure for “spiked” and “unspiked” dissolver solutions were used in TIMS measurements for the first time in our laboratory. In TIMS analysis, isotopic abundances of uranium, plutonium and neodymium fractions obtained from HPLC separation procedure on an “unspiked” fuel sample were measured. For the determination of U, Pu and Nd by isotopic dilution mass spectrometric technique (IDMS), known quantities of tracers enriched in 238U, 240Pu and 142Nd were added to the pre-weighed dissolver solution and subjected to HPLC separation procedures. The isotope ratios viz. 142Nd/(145Nd +146Nd), 238U/233U and 240Pu/239Pu in the pertinent “spiked” fractions were subsequently measured by TIMS. The spikes were pre-standardized in our laboratory employing reverse isotopic dilution technique against the standard solutions available in our laboratory (for 238U, 239Pu and 142Nd, standard solutions of 233U, 239Pu (of higher abundance than in the sample) and 150Nd were employed as spikes). The burn-up values from duplicate spiking experiments were computed based on the summation of 145Nd + 146Nd. The concentrations of neodymium, uranium and plutonium were also measured using HPLC with post-column derivatisation technique using aresenazo(III) as the post-column reagent. The atom % burn-up computed from HPLC and TIMS techniques were in good agreement.