Design of a fuel element for a lead-cooled fast reactor

The options of a lead-cooled fast reactor (LFR) of the fourth generation (GEN-IV) reactor with the electric power of 600 MW are investigated in the ELSY Project. The fuel selection, design and optimization are important steps of the project. Three types of fuel are considered as candidates: highly enriched Pu-U mixed oxide (MOX) fuel for the first core, the MOX containing between 2.5% and 5.0% of the minor actinides (MA) for next core and Pu-U-MA nitride fuel as an advanced option. Reference fuel rods with claddings made of T91 ferrite-martensitic steel and two alternative fuel assembly designs (one uses a closed hexagonal wrapper and the other is an open square variant without wrapper) have been assessed. This study focuses on the core variant with the closed hexagonal fuel assemblies. Based on the neutronic parameters provided by Monte-Carlo modeling with MCNP5 and ALEPH codes, simulations have been carried out to assess the long-term thermal-mechanical behaviour of the hottest fuel rods. A modified version of the fuel performance code FEMAXI-SCK-1, adapted for fast neutron spectrum, new fuels, cladding materials and coolant, was utilized for these calculations. The obtained results show that the fuel rods can withstand more than four effective full power years under the normal operation conditions without pellet-cladding mechanical interaction (PCMI). In a variant with solid fuel pellets, a mild PCMI can appear during the fifth year, however, it remains at an acceptable level up to the end of operation when the peak fuel pellet burnup ∼80 MW d kg⁻¹ of heavy metal (HM) and the maximum clad damage of about 82 displacements per atom (dpa) are reached. Annular pellets permit to delay PCMI for about 1 year. Based on the results of this simulation, further steps are envisioned for the optimization of the fuel rod design, aiming at achieving the fuel burnup of 100 MW d kg⁻¹ of HM.     

Short-term irradiation behavior of minor actinide doped uranium plutonium mixed oxide fuels irradiated in an experimental fast reactor

A fuel irradiation program is being conducted using the experimental fast reactor ‘Joyo’. Two short-term irradiation tests in the program were completed in 2006 using a uranium and plutonium mixed oxide fuel which contains minor actinides (MA-MOX fuel). The objective of the tests is the investigation of early thermal behavior of MA-MOX fuel such as fuel restructuring and redistribution of minor actinides. Three fuel pins which contained MA-MOX: 2% neptunium and 2% americium doped uranium plutonium mixed oxide (Am,Pu,Np,U)O_2−x fuel were supplied for testing. The first test was conducted with high-linear heating rate of approximately 430 W cm⁻¹ for only 10 min. After the first test, one fuel pin was removed for examinations. Then the second test was conducted with the remaining two pins at nearly the same linear power for 24 h. In these tests, two oxygen-to-metal molar ratios were used for fuel pellets as a test parameter. Non-destructive and destructive post-irradiation examinations results are discussed with early on the behavior of the fuel during irradiation.     

Microstructure and elemental distribution of americium-containing uranium plutonium mixed oxide fuel under a short-term irradiation test in a fast reactor

In order to investigate the effect of americium addition in MOX fuel on the irradiation behavior, the ‘Am-1’ program is being conducted in the experimental fast reactor Joyo. The Am-1 program consists of two short-term irradiation tests of 10 min and 24 h irradiations and a steady-state irradiation test. The short-term irradiation tests were successfully completed and the post irradiation examinations (PIEs) are in progress. This paper reports on the results of PIEs for Am-containing MOX fuel irradiated for 10 min. MOX fuel pellets containing 3% or 5% Am were fabricated in a shielded air-tight hot cell using a remote handling technique. The oxygen to metal ratio (O/M) of these fuel pellets was 1.98. They were irradiated at peak linear heating rate of about 43 kW m⁻¹. Focus was being placed on migration behavior of Am during the irradiation. The ceramography results showed that structural changes such as lenticular pores and a central void occurred early, within the brief 10 min of irradiation. The results of electron probe microanalysis revealed that the concentration of Am increased in the vicinity of the central void.     

Radial redistribution of actinides in irradiated FR-MOX fuels

The redistributions of neptunium, plutonium and americium during two kinds of short-term irradiation tests for 10 min and 24 h at high linear heating rate around 430 W cm⁻¹ were studied in the uranium and plutonium mixed oxide fuel containing Am and/or Np. It was found in the irradiation test for 24 h that the concentrations of Pu and Am increased toward the central void, but there was no change in the concentration of Np. The obtained experimental redistributions of Am and Pu were analyzed, based on both pore migration and thermal diffusion models. As a result, the calculated redistributions of Pu and Am showed good agreements with the experimentally obtained ones.     

Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO₂ fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor k_inf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO₂ fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO₂, UO₂ with 4.0 vol.% BeO, and UO₂ with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO₂ fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the k_inf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO₂.     

Behavior of Si impurity in Np-Am-MOX fuel irradiated in the experimental fast reactor Joyo

The irradiation behavior of uranium-plutonium mixed oxide fuels containing a large amount of silicon impurity was examined by post-irradiation examination. Influences of Si impurity on fuel restructuring and cladding attack were investigated in detail. Si impurity, along with Am, Pu and O were transported by spherical pores and cylindrical tubular pores to the fuel center during fuel restructuring of the Np-Am-MOX fuel, where a eutectic reaction of fuel and Si-rich inclusions occurred. After fuel restructuring of the Np-Am-MOX fuel, Si-rich inclusions without fuel constituents were agglomerated at fuel crack openings where shallow attacks on the inner wall of the cladding were seen. Such shallow attacks on the inner wall of the cladding were likewise observed near the location of fuel cracks in long-term steady-state irradiated MOX fuels. Evidence of these shallow attacks on the inner wall of the cladding remained after fuel restructuring in normal MOX fuel. However, grain boundary corrosion of the cladding inner wall at the opening of the fuel cracks was selective and was marked in MOX fuel at higher oxygen potential by the release of reactive fission products such as Cs and Te in comparison with other regions of cladding wall.     

Oxygen potential of hypo-stoichiometric Lu-doped UO2

Oxygen potentials of hypo-stoichiometric Lu-doped UO₂, (U_0.80Lu_0.20)O_2−x, were experimentally investigated by thermogravimetric analysis using H₂O/H₂ gas equilibria at 1173, 1273 and 1473 K. The oxygen potentials of (U,Lu)O_2−x were higher than those of other forms of rare earth-doped UO₂, specifically (U,Nd)O_2−x, (U,Gd)O_2−x, and (U,Er)O_2−x. Slope analyses for plots of oxygen potential versus deviation from stoichiometry indicated that (U_0.80Lu_0.20)O_2−x had a similar defect structure to that of the other forms of rare earth-doped UO₂. A relationship between the effective ionic radii and oxygen potentials was found for the hypo-stoichiometric rare earth-doped UO₂.     

New concept of designing Pu and MA containing fuel for fast reactors

New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO₂ powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO₂ powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.     

Equilibrium and nonequilibrium molecular dynamics simulations of heat conduction in uranium oxide and mixed uranium-plutonium oxide

The thermal conductivity of nuclear fuels such as UO_2+x and (U,Pu)O_2−x has been calculated by the molecular dynamics (MD) simulation in terms of oxygen stoichiometric parameter x, temperature and Pu content. In the present study, the MD calculations were carried out in both equilibrium (EMD) and nonequilibrium (NEMD) systems. In the EMD simulation, the thermal conductivity was defined as the time-integral of the correlation function of heat fluxes according to the Green-Kubo relationship. Meanwhile, in the homogeneous NEMD, it was given by the ratio of the time-averaged heat flux to the perturbed external force subjected to each particle in the simulated cell. NEMD, as compared with EMD, gave somewhat precise results efficiently. Furthermore, both MD calculations showed that the thermal conductivity of these oxide fuels decreased with increase of temperature and defects, i.e. excess oxygen or vacancy, and was rather insensitive to Pu content for the stoichiometric fuel.     

Fusion-Fission Transmutation Scheme—Efficient destruction of nuclear waste

A fusion-assisted transmutation system for the destruction of transuranic nuclear waste is developed by combining a subcritical fusion-fission hybrid assembly uniquely equipped to burn the worst thermal nonfissile transuranic isotopes with a new fuel cycle that uses cheaper light water reactors for most of the transmutation. The center piece of this fuel cycle, the high power density compact fusion neutron source (100 MW, outer radius <3 m), is made possible by a new divertor with a heat-handling capacity five times that of the standard alternative. The number of hybrids needed to destroy a given amount of waste is an order of magnitude below the corresponding number of critical fast-spectrum reactors (FRs) as the latter cannot fully exploit the new fuel cycle. Also, the time needed for 99% transuranic waste destruction reduces from centuries (with FR) to decades.     

Monte Carlo simulation of the Greek Research Reactor neutron irradiation facilities

A Monte Carlo simulation of the Greek Research Reactor was carried out using MCNP-4C2 code and continuous energy cross-section data from ENDF/B-VI library. A detailed model of the reactor core was employed including standard and control fuel assemblies, reflectors and irradiation devices. The model predicted neutron flux distributions within the core in good agreement with calculations performed using the deterministic code CITATION and measurements using activation foils. The model is used for the prediction of the neutron field characteristics at the reactor irradiation devices and enables the design and evaluation of experiments involving material irradiations.     

Spark Plasma Sintering of simulated radioisotope materials within tungsten cermets

A Spark Plasma Sintering (SPS) furnace was used to produce ceramic-metallic sinters (cermets) containing a simulated loading of radioisotope materials. CeO₂ was used to simulate loadings of PuO₂, UO₂ or AmO₂ within tungsten-based cermets due to the similar kinetic properties of these materials, in particular the respective melting points and Gibbs free energies. The work presented demonstrates the capability and suitability of the SPS process for the production of radioisotope encapsulates for nuclear fuels and other applications (including waste disposal and radioisotope power and heat source fabrication) where the mechanical capture of radioisotope materials is required.     

Fabrication of uniform ZrC coating layer for the coated fuel particle of the very high temperature reactor

Fuel for the very high temperature reactor is required to be used under severer irradiation conditions and higher operational reactor temperatures than those of present high temperature gas cooled reactors. Japan Atomic Energy Agency has developed zirconium carbide (ZrC)-coated fuel particles previously in laboratory scale which are expected to maintain their integrity at higher temperatures and burnup conditions than conventional silicon carbide-coated fuel particles. As one of the important R&D items, ZrC coating process development has been started in the year 2004 to determine the coating conditions to fabricate uniform structure of ZrC layers by using a new large-scale coater up to 0.2 kg batch. It was thought that excess carbon formed in the ZrC layer under the oscillation of coating temperature would cause non-uniformity of the ZrC layer. Finally, uniform ZrC coating layer has been fabricated successfully by adjusting the time constant of the coater and keeping the coating temperature at around 1400 °C.     

A method for the quantification of total xenon concentration in irradiated nuclear fuel with SIMS and EPMA

A new method for the quantitative determination of the total xenon concentration in irradiated nuclear fuel is presented. The SIMS measurement of xenon enables the detection of the gas filling bubbles which are not detected with EPMA. The quantification is achieved using the EPMA data as reference at position where no or nearly no bubbles are detected. A new approach using the complementary information given by EPMA, SEM and SIMS is proposed, it opens new horizons for the characterisation of fission gases in irradiated nuclear fuel.     

Neutronic design of a sub-critical, source driven reactor at the RA-8 facility

We present a LEU-ADS design based on an existing Argentine experimental facility, the RA-8 pool type zero power reactor. The versatility of this reactor allows measurement of different core configurations using different fuel enrichment, burnable poison rods, water perturbations, different control rods types in critical or subcritical configurations with an external source.To assess the feasibility of the LEU-ADS, multiplication factors, kinetic parameters, spectra, and time flux evolution were computed. Two external sources were considered: an isotopic source, and a D-D pulsed neutron source.Parameters for different core configurations were calculated, and the feasibility of using continuous and pulsed neutron sources was verified.     

Study of atomic clusters in neutron irradiated reactor pressure vessel surveillance samples by extended X-ray absorption fine structure spectroscopy

Copper and nickel impurities in nuclear reactor pressure vessel (RPV) steel can form nano-clusters, which have a strong impact on the ductile-brittle transition temperature of the material. Thus, for control purposes and simulation of long irradiation times, surveillance samples are submitted to enhanced neutron irradiation. In this work, surveillance samples from a Swiss nuclear power plant were investigated by extended X-ray absorption fine structure spectroscopy (EXAFS). The density of Cu and Ni atoms determined in the first and second shells around the absorber is affected by the irradiation and temperature. The comparison of the EXAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the EXAFS analysis reveals local irradiation damage in the form of vacancy fractions, which can be determined with a precision of ∼5%. There are indications that the formation of Cu and Ni clusters differs significantly.     

Method based on isotope ratio mass spectrometry for evaluation of carbon activation in the reactor graphite

The general idea of this work is to introduce an evaluation method to restore the irradiation parameters of graphite or other carbonaceous materials using experimental and modelling results of ¹³C generation in the irradiated material. The method is based on coupling of stable isotope ratio mass spectrometry and computer modelling of the reactor core to evaluate the realistic characteristics of the reactor core such as the neutron fluence in any position of the reactor graphite stack or other graphite constructions.The generation of carbon isotopes ¹³C and ¹⁴C in the irradiated graphite of the RBMK-1500 reactor has been estimated by modelling of the reactor core with computer codes MCNPX and CINDER90. Good agreement of simulated and measured Δ¹³C/¹²C values in graphite of the central part of the reactor core indicates that the neutron flux (1.40 × 10¹⁴ n/cm² s) is modelled accurately in the graphite sleeve of the fuel channel. The simulated activity of ¹⁴C is compared with the one measured by the β spectrometry technique. Results indicate that production of ¹⁴C from ¹⁴N in the RBMK-1500 reactor is considerable and has to be taken into account in order to make proper evaluation of ¹⁴C activity. Measured ¹⁴C specific activity values correspond to 15 ± 4 ppm impurity of ¹⁴N in graphite samples from the RBMK-1500 reactor core.     

Particulars of neutron-physical calculations of sodium-cooled fast reactors with mixed oxide fuel

The neutron-physical characteristics of reactor systems with a fast spectrum, sodium coolant, and uraniumplutonium fuel load have been analyzed on the basis of computational studies of the BFS-62-3A critical assembly and a BN-600 hybrid core with mixed oxide fuel. The large differences in the spectra in an expanded thermal range to 1 keV for the central and peripheral regions with uranium oxide and mixed oxide fuel show that spatially differentiated fission and absorption cross sections must be used for the main uranium and plutonium isotopes in neutron-physical calculations.     

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.