Elements of comparison between different inert matrix fuels as regards plutonium utilisation and safety coefficients

The current international trend is to focus towards the utilization of plutonium. The use of composite fuels in inert matrix (U-free) is a potentially efficient solution to this problem. This document deals with the cermet fuels, selected for their excellent behaviour under irradiation and their high thermal conductivity. The emphasis is placed on the study of kinetic coefficients. Comparisons are performed with other solutions that use other composite fuels, especially the Solid Solutions and ROXs. As core control requires a heterogeneous assembly, an assembly whose characteristics are compared to the APA reference is proposed.     

Inert matrix and thoria fuels for plutonium elimination

In our previous studies we analysed the plutonium burning performance of various kinds of fuel, based on mixing plutonium oxide with thorium oxide (TOX), or with inert matrix (IMF), or with inert matrix plus a limited addition (doping) of thorium oxide (TD-IMF). The present paper includes the neutronic analysis of a metal-based fuel variant and of fissile material recycling in TOX fuels. If the recycled fuel is topped with weapons grade plutonium (WG-Pu) as fissile material, it is possible to sustain indefinitely a closed cycle.     

Radiation damage and simulated fission product effects on the properties of inert matrix materials

Candidate inert matrix materials for actinide transmutation (MgAl₂O₄, CeO₂) or immobilization (ZrSiO₄) containing ²⁴¹Am were characterized. The currently most considered material, ZrO₂, was produced, with La₂O₃ as stand-in for Am, and with and without simulated fission products to investigate burnup effects. The oxygen potential was measured using an EMF cell. The accumulation of radiation damage due to Am decay was investigated by periodically measuring lattice parameters and hardness. Sequential leaching tests in deionized water, aimed at correlating the leaching behaviour of Am and of the matrix with radiation damage, showed significant release of Am and of some matrix components.     

Inert matrix fuel for the utilisation of plutonium

Utilising fuel resources responsibly, reducing waste volume and emissions as well as conflict potentials within the international community (non-proliferation, energy demand) are among the principles for the judgment of sustainable development. Utilising and burning plutonium in a light water reactor has been shown to be feasible for the disposition of the large amount of excess plutonium produced in today's power reactors and resulting from the disarmament efforts of the super powers. With regard to material technology aspects, efforts have concentrated on the evaluation of fabrication feasibility and on the determination of the physicochemical properties of a single phase zirconium/erbium/plutonium oxide material stabilised as a cubic solution by yttrium for plutonium and minor actinide incineration and transmutation. Due to the absence of uranium as origin for plutonium build-up, such a nuclear fuel is called Inert Matrix Fuel. Irradiation testing with a dedicated experiment in the material test reactor in Halden, Norway, is underway within the framework of the OECD-Halden Reactor Project.     

Neutronic and safety aspects of inert matrix fuel utilization in fast reactors for plutonium and minor actinides transmutation

The solving of ecological problems of future nuclear power is connected with the solving of long-lived radioactive waste utilization problems. It concerns primarily plutonium and minor actinides (MAs), accumulated in the spent fuel of nuclear reactors. One of the ways this can be solved is to use a fast reactor with uranium-free or inert matrix fuel (IMF). The physics of this type of reactor was widely investigated during last year for BN-800 reactors. The solution of the most important problems was: a decrease in non-uniformity of power distribution and an increase of the Doppler effect. The next stage of such core investigations is an evaluation of self-protection to beyond design accidents. Preliminary results show a high safety level of BN-800 reactors with IMF in the event of unprotected loss of coolant flow (ULOF) accident.     

Integral measurements with a plutonium inert matrix fuel rod in a heterogeneous light water reactor lattice

As part of PSI's validatory efforts for neutronics calculations pertaining to inert matrix fuel (IMF) deployment in light water reactors, first-of-their-kind integral measurements have been carried out at the PROTEUS facility employing a specially fabricated Pu-Er-Zr IMF rod in a heterogeneous (boiling water reactor) test lattice. Analogous experiments have also been conducted with MOX and “dummy” IMF rods, providing the basis for a systematic comparison of experimental and calculational results.     

A numerical neutronics benchmark study for inert matrix plutonium fuels in uranium dioxide and mixed plutonium-uranium dioxide environments

A numerical benchmark exercise has been under way for comparing the results of different calculational methods/data sets used for the analysis of light water reactor (LWR) configurations employing Pu inert matrix fuels (IMFs). The first phase of the exercise was devoted to infinite arrays of identical IMF cells. The main feature investigated in the second phase has been the influence of the neutron spectra in UO₂ and MOX cores on individual IMF cells. Phase 3 is concerned with the more realistic situation of an IMF assembly surrounded by UO₂ assemblies. Significant discrepancies have been observed for power peaking effects and delayed neutron parameters in Phase 2. In Phase 3, neutron balance differences for the IMF, particularly at EOL, are found to be significantly larger than were observed in Phase 1.     

X-ray diffraction analysis and data interpretation of stabilized zirconia inert matrix fuel doped with plutonium

Yttria stabilized zirconia doped with erbia and plutonia has been selected as an inert matrix fuel (IMF) at PSI in order to destroy fissile plutonium in the form of a uranium-free fuel in an effective way. The crystallographic structure (lattice parameters) of cubic zirconia strongly depends on the choice of the stabilizer and other dopants i.e. burnable poisons or fissile material. An extensive study of X-ray diffraction measurements was performed on zirconia samples containing different amounts of additives with the aim to observe lattice parameter and crystallite size changes in the IMF. A semi-quantitative model already available in literature was used and adapted to predict the “theoretical” lattice parameters of IMF with plutonia. The results show a good agreement of theory and experiment. Furthermore, for the first time the structure of active IMF based on zirconia has been investigated and been compared to the X-ray diffraction patterns of undoped zirconia. As a consequence, it is now possible to predict lattice parameters and final densities of IMF with varying compositions, and a good control of the sample dimensions during the fabrication can be guaranteed.     

Radiotoxicity hazard of inert matrix fuels after burning minor actinides in light water reactors

Ingestion radiotoxicity hazard index of inert matrix spent fuels are investigated after burning minor actinide (MA) isotopes in LWRs and compared with the hazard index of MOX and MA burning MOX (MOX+MA) spent fuels. As U-free fuels, ROX: (PuO₂+ZrO₂) and TOX: (PuO₂+ThO₂), are considered, in which MA's are added as oxides. The radiotoxicity hazard index of ROX+MA spent fuel is less than that of TOX+MA and MOX+MA spent fuels due to the lower density of actinides in spent fuel. Some of cooling years the toxic yield of ROX+MA spent fuel is even less than that of MOX spent fuel, if the initial loaded MA in ROX is about 0.5 at %.     

Design and fabrication aspects of a plutonium-incineration experiment using inert matrices in a “Once-Through-Then-Out” mode

In the plutonium incineration experiment, named ‘Once-Through-Then-Out’ (OTTO), that is being prepared by JAERI, PSI and NRG, the use of highly stable inert matrices will be examined. The inert matrices MgAl₂O₄ spinel and ZrO₂ are insoluble in nitric acid and are considered as good storage media for final disposal. These inert matrices will be used in this experiment, which is representative for an OTTO scenario. A total of 7 Pu-containing targets were prepared for an irradiation in the High Flux Reactor in Petten. The objective of the irradiation is to reach a very high Pu-burnup. The main parameters to be studied are stability under irradiation, swelling, fission gas release and chemical interactions in the fuel. Four targets will be equipped with thermocouples for on-line monitoring of central temperature. Four of the targets contain MgAl₂O₄ as an inert matrix, 2 targets contain ZrO₂ and one target contains mixed-oxide (MOX) fuel for reference purposes. The fissile plutonium concentration is 0.32–0.44 g cm⁻³. Both particle-dispersed fuel and homogeneous dispersions were fabricated in order to test the effect of the size of the fissile inclusions. The design of the experiment and the fabrication of the samples are discussed.     

Composite materials and systems as alternative inert matrix fuel to dispose civil or weapon grade plutonium in light water reactors.

To complete the IMF cercer studies on the problem of Pu utilization in LWRs, a cermet fuel approach is presented. The advantages of cermet fuel are associated with high heat conductivity, ability to retain the fission products and a well-developed fabrication process. Attractive possibilities for the creation of new cermet fuels and cermet fuel elements are also presented. R&D activity aimed at the development of cermet fuel element with PuO₂-Zr composite was undertaken. As a result of this activity comparative analysis of thermodynamic calculations for UO₂-Zr and PuO₂-Zr composites was carried out, as well as an assessment of Pu loading and preliminary thermal calculations. As a consequence, it was concluded that the PuO₂-Zr cermet system could be considered as a possible variant of new cermet fuel and cermet fuel element for Pu burning in LWRs.     

Analysis of porous features in zirconia based inert matrix, impact on the material qualification

Pore size distribution, porosities and consequently derived densities are key parameters for the qualification of inert matrix fuel (IMF). Porous features must be determined accurately for assessing the in-pile behaviour of the fuel. The analytical methodologies to measure porosities are revisited. The paper discusses in a comprehensive way the results obtained for simulated or plutonium loaded inert matrix fuel i.e. erbia and ceria or plutonia doped yttria stabilised zirconia, by applying thermodynamic or instrumental approaches. The n-D results (open/close porosities) gained by applying invasive/non-invasive, microscopic/macroscopic, neutron/photon and reflection/diffraction/transmission methods with fluid intrusion or not, are given and discussed for the zirconia based cubic solid solution.     

Current status of researches on the plutonium rock-like oxide fuel and its burning in light water reactors

Intention of the ROX-LWR system research is to provide an option for utilization or disposition of surplus plutonium. Researches on inert matrix materials and irradiation performance shows that the most favorable candidate for the ROX fuel is a particle dispersed fuel where small particles consisted of yttria stabilized zirconia, PuO₂ and some additives are homogeneously dispersed in spinel matrix. Reactor safety analyses show that the ROX fueled PWR core has nearly the same performability as the existing UO₂ fueled PWR under both reactivity initiated accidents and loss of coolant accidents.     

Erbium: alternative poison? Stabilisation additive? What future?

Erbium was proposed as alternative poison to gadolinium at a very early stage. The potential interest of this poison compared to gadolinium is that it presents a relatively low (¹⁶⁷Er) absorption cross section in the thermal range and a non-negligible resonance integral that leads to a relatively slow consumption kinetic rather adapted to long or even very long cycles. The poisoning mode with this element, when homogeneous in low concentration (< 3 %), does not downgrade the power distribution, on the one hand, as the absorption is low and spatially homogeneous, and the thermal conductivity, on the other hand, as the addition in the fuel oxide is in low quantity. A review of knowledge acquired as regards erbium, from the 1960s to now, is presented.     

Proposal of a plutonium burner system based on HTGR with high proliferation resistance

An innovative plutonium burner concept based on high temperature gas cooled reactor (HTGR) technology, “Clean Burn”, is proposed by Japan Atomic Energy Agency (JAEA). That is expected to be as an effective and safe method to consume surplus plutonium accumulated in Japan. A similar concept proposed by General Atomics (GA), Deep Burn, cannot be introduced to Japan because of its adopting highly enriched plutonium, which shall infringe on a Japanese nuclear nonproliferation policy according to Japan–US reprocessing negotiation. The Clean Burn concept can avoid this problem by employing an inert matrix fuel (IMF) and a tightly coupled fuel reprocessing and fabrication plants. Both features make it impossible to extract plutonium alone out of the fabrication process and its outcomes. As a result, the Clean Burn can use surplus plutonium as a fuel without mixing it with uranium matrix. Thus, surplus plutonium alone will be incinerated effectively, while generation of plutonium from the uranium matrix is avoided. High neutronic performance, i.e., achievement of burn-up of about 500 GWd/t and consumption ratio of plutonium-239 reaching to about 95%, is also assessed. Furthermore, reactivity defect caused by the inert matrix is found to be negligible. It is concluded that the Clean Burn concept is a useful option to incinerate plutonium with high proliferation resistance.     

Measurement of the mechanical properties of thermally-shocked zirconia-based simulated inert matrix fuel

The Vickers micro-hardness (H_V) was measured by an indentation technique of simulated ZrO₂-based Inert Matrix Fuel (IMF) material with a composition of Er_0.07Y_0.10Ce_0.15Zr_0.68O_1.915 in two different densities on sintered specimens and specimens thermally shocked with the quenching temperature differences (ΔTs) between 473 and 1673 K and compared with those of simulated MOX, namely, U_0.92Ce_0.08O₂. The H_V values obtained for two IMF materials were found higher, ranging from 6.37 GPa to about 7.84 GPa, depending on ΔT and the sintered density, than those obtained for the simulated MOX which are quasi-constant in the same range of ΔT with a mean value of 6.37 GPa. The fracture toughness (K_IC) was calculated from the measured H_V and the crack length, and it was found to exhibit a slight increase with increasing ΔT, ranging between 1.4 and 2.0 MPa m^1/2, while that of simulated MOX specimen is ranging between 0.8 and 1.1 MPa m^1/2. The thermally shocked specimens were observed with an optical microscope and analyzed in terms of microstructural changes and cracking patterns.     

Boron carbide as a potential inert matrix: an evaluation

Materials such as carbides, borides or silicides offer good prospects for use as inert matrices for actinide burning. We present here an evaluation of the properties of one of these materials, boron carbide B₄C. To test its ability to host other elements, we have prepared and characterised ceramic-ceramic composites in which the added phase is simulated by hafnium diboride: we have then shown that such materials have improved thermo-mechanical properties. Concerning actinide disposition, we present preliminary results concerning the preparation and the characterisation of composites containing cerium compounds as the added phase.     

Release of cesium from zirconia inert matrix

The release upon annealing or irradiation at high temperature of fission products (Cs) from zirconia, a candidate for inert matrix, is studied by Rutherford backscattering experiments. The work is focused on the understanding of the influence of various parameters (mainly the atomic concentration of foreign species and the damage created by an external irradiation) on the diffusion and release of Cs. The results reveal that the Cs mobility is strongly enhanced when the impurity concentration exceeds a threshold of the order of a few atomic per cent. Irradiation with medium-energy heavy ions is shown to cause a further increase of the Cs mobility.     

Fuel cycle strategies and plutonium management in Europe

An overview of current nuclear power generation and fuel cycle strategies in Europe is presented, with an emphasis on options for the management of separated plutonium in the medium to long term. Countries which have opted for reprocessing of spent fuel have had to contend with increasing inventories of separated plutonium. Of the various potential options for utilisation or disposition of these stockpiles, only light water reactor (LWR) mixed-oxide (MOX) fuel programmes are sufficiently technologically mature to be fully operational in several European countries at present. Such reprocessing-recycling programmes allow for a stabilisation of the overall separated plutonium stocks, but not for a significant reduction in the stockpile. Moreover, the quality of recycled plutonium decreases at each potential step of re-irradiation. Therefore, optimised or new ways of managing the plutonium stocks in the medium to long term are required. In the present overview we consider the most promising options for reactor utilisation of plutonium in both near-term future reactor and Generation IV systems.     

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from 10² to 10⁵ years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.