Transmutation Fuels Research Articles

Phase relations and lattice parameter trends in Nd3+−substituted UO2 system under reducing and oxidizing conditions

While Am-substituted UO2 solid solutions have been considered as advanced fuels for heterogeneous transmutation of americium, U-substituted Am2O3 with cubic structure has been envisaged as a promising candidate for radioisotope heat unit and radioisotope thermoelectric generator. Considering Nd2O3 as a surrogate for Am2O3, a series of Nd3+-substituted UO2 samples with the general formula U1-xNdxO2±δ (0 ≤ x ≤ 1, δ ≥ 0) have been investigated for phase relations under reducing and oxidizing conditions. In this direction, samples have been synthesized by a solid-state route and thoroughly characterized by X-ray diffraction and Raman spectroscopy. XRD studies revealed the formation of fluorite-type solid solution at x ≤ 0.7 and a biphasic phase field consisting of C-type (U,Nd)O2−δ phase and A-type Nd2O3 at x ≥ 0.8 in reducing conditions. In the fluorite-type regime, due to the aliovalent Nd3+-substitution in UO2, the charge neutrality was achieved primarily through oxygen vacancy creation as evident from lattice parameter trends as well as Raman spectroscopic investigation that exhibit Raman band due to oxygen vacancies. The reduced samples have a high tendency to get oxidized even during storage under ambient conditions leading to a decrease in lattice parameters, which can be attributed to the formation of the higher oxidation state of uranium (U5+/U6+) following a charge compensation mechanism. Synthesis in the oxidizing conditions yielded a biphasic phase field at x ≤ 0.2, a single-phase fluorite-type phase field at 0.3 ≤ x ≤ 0.7, a biphasic mixture of fluorite-type (U,Nd)O2−δ and trigonal Nd6UO12 phases at x = 0.8 and another biphasic mixture of Nd6UO12 and Nd2O3 phases at x = 0.9. The transition of reduced single-phase fluorite-type solid solution at x ≤ 0.2 to a biphasic mixture upon oxidation has been probed by a high-temperature XRD.

Current state of partitioning and transmutation studies for advanced nuclear fuel cycles

This paper aims at giving a general overview of the current state of partitioning and transmutation (P&T) studies for advanced nuclear fuel cycles across the world. P&T usually refers to the entire fuel cycle steps linked with minor actinides transmutation, from separation of the minor actinides from the spent fuel to manufacturing of new fuels suited for transmutation, irradiation, handling and reprocessing or disposal of the spent transmutation fuels. The feasibility of most of these steps has been demonstrated in the laboratory scale since the beginning of the nuclear era, however the physical-chemical and radiological properties of minor actinides make their handling considerably more complicated than standard UOX or MOX fuel. As such, research on this topic is still ongoing around the world.Due to the shutdown of various experimental installations since the 90′ (EBR-II in the US in 1994, PFR in the UK in the same year, Joyo in Japan in 2007, Phénix and Osiris in France in 2009 and 2016 respectively), few irradiation experiments have been carried out in the past decade. The MARIOS and DIAMINO experiments were carried out in the HFR and OSIRIS reactor and showed that the irradiation of fuel with a micro-structure optimized for helium release (and thus limited swelling) was achievable. The irradiation of metallic and oxide containing minor actinides and/or rare earths was carried out in the ATR reactor in the US. The good behaviour of oxide fuel containing minor actinides was demonstrated, although pin failures occurred with the metallic fuel. Furthermore, Post-Irradiation Examinations of various irradiation experiments have been carried out, the most salient one being the METAPHIX experiment in Phenix, which demonstrated the safe behaviour under irradiation of the base alloy considered in the experiment.Nevertheless, encouraging results in the field of americium separation from the spent fuel have been obtained, the main one being the successful demonstration of the ExAm process to isolate americium from a PUREX raffinate. Furthermore, progress has been made on the manufacturing of UAmO2 targets during the preparation of DIAMINO and MARIOS experiments. Finally, interest in minor actinides transmutation has started to rise again, notably linked with the current trend in molten salt reactor, which are suitable for minor actinides transmutation.

Remote injection casting process with reduced material loss for fabrication of metallic fuels

The remote injection casting process was developed using a high-capacity casting system in a mockup facility to secure the key technology for the fabrication of transmutation metallic fuels in hot cells. Seventy-eight fuel slugs fabricated simultaneously through remote injection casting were confirmed to satisfy the fabrication specification. The reaction characteristics of the coating candidates for the reusable crucible were investigated by using sessile drop tests with U-10wt%Zr-5wt%RE at 1450 °C for 1 h and LaYO3 showed the smallest interaction width. The improvement in the adhesion of Y2O3 coatings on the inner walls of the quartz mold was effective in reducing the reaction between the melt and the mold. The mechanical and chemical cleaning processes were investigated to remove the surface contamination layer of the melt residue for recycling. Both cleaning methods were effective with the mechanical treatment favored owing to the operations in hot cells. The remote injection casting process equipped with fuel loss control components developed in this work will fabricate transmutation fuels in hot cells.

Preliminary study of fabrication technology for transmutation fuels using surrogate alloy

A preliminary study on the fabrication of transmutation fuels was carried out using an injection casting technique. Melts of binary U–Zr, a surrogate alloy for transmutation fuels, with high Zr contents (15, 20, and 25 wt%) were mixed via electromagnetic stirring, and fuel slugs were successfully cast to the full length of the mold without macrodefects. From chemical composition analysis by ICP-AES, the Zr content of the U–Zr slugs matched well with the target composition, to within 5% difference. The microstructure examined by SEM showed that with an increase in the Zr concentration, the lamellar structure was reduced, and Zr-rich inclusions were uniformly distributed over the fuel slugs.

Melting behaviour of uranium-americium mixed oxides under different atmospheres

In the context of a comprehensive campaign for the characterisation of transmutation fuels for next generation nuclear reactors, the melting behaviour of mixed uranium-americium dioxides has been experimentally studied for the first time by laser heating, for Am concentrations up to 70 mol. % under different types of atmospheres. Extensive post-melting material characterisations were then performed by X-ray absorption spectroscopy and electron microscopy. The melting temperatures observed for the various compositions follow a markedly different trend depending on the experimental atmosphere. Uranium-rich samples melt at temperatures significantly lower (around 2700 K) when they are laser-heated in a strongly oxidizing atmosphere compressed air at (0.300 ± 0.005) MPa, compared to the melting points (beyond 3000 K) registered for the same compositions in an inert environment (pressurised Ar). This behaviour has been interpreted on the basis of the strong oxidation of such samples in air, leading to lower-melting temperatures. Thus, the melting temperature trend observed in air is characterized, in the purely pseudo-binary dioxide plane, by an apparent maximum melting temperature around 2850 K for 0.3 < x(AmO2) < 0.5. The melting points measured under inert atmosphere uniformly decrease with increasing americium content, displaying an approximately ideal solution behaviour if a melting point around 2386 K is assumed for pure AmO2. In reality, it will be shown that the (U, Am)-oxide system can only be rigorously described in the ternary U-Am-O phase diagram, rather than the UO2-AmO2 pseudo-binary, due to the aforementioned over-oxidation effect in air. Indeed, general departures from the oxygen stoichiometry (Oxygen/Metal ratios ≠ 2.0) have been highlighted by the X-ray Absorption Spectroscopy (XAS). Finally, to help interpret the experimental results, thermodynamic computations based on the CALPHAD method will be presented.

The microstructure and thermodynamic behavior of as-cast U-24Pu-15Zr: Unexpected results and recommendations for U-Pu-Zr fuel research methodology

Minor actinide transmutation fuels, of which U-Pu-Zr is one of the most promising, have been the subject of renewed interest for fast reactor applications. Unfortunately, we lack the understanding necessary to make quantitative, mechanistic predictions about the complex phase behaviors exhibited by U-Pu-Zr. This prevents the efficient development and implementation of U-Pu-Zr fuels. Herein, we use state of the art experimental and mesoscale simulation techniques to examine and predict the behavior of U-24Pu-15Zr (weight percent). Experimental and simulated microstructural and phase stability results are compared to one another and to accepted reference data. Experiments revealed a heterogeneous microstructure composed of αZr, ZrO2, δ, βPu, and ζ at room temperature. The unexpectedly high ratio of βPu to ζ is believed to arise from non-equilibrium and surface effects, as both passivation and Pu and Zr segregation were observed in the samples. A modeling technique based on the overall bulk free energy is devised to estimate the extent of a microstructure's departure from equilibrium. Agreement between the datasets is used to justify a set of recommendations for the further study of U-Pu-Zr fuels. These address deficiencies in our fundamental knowledge, fuel fabrication and handling techniques, characterization methods and procedures, modeling capability, and use of coupled experiments and mesoscale simulations.

Isolation of high purity americium metal via distillation

Pure americium metal is a crucial component for the fabrication of transmutation fuels. Unfortunately, americium in pure metal form is not available; however, a number of mixed metals and mixed oxides that include americium are available. In this manuscript a method is described to obtain high purity americium metal from a mixture of americium and neptunium metals with lead impurity via distillation.

Target fuels for plutonium and minor actinide transmutation in pressurized water reactors

The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, containing approximately 95wt% uranium, 1wt% plutonium, and 4wt% fission products and transuranic elements. Fast reactors are a preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20years indicates a need for a nearer-term solution. This study considers a method for plutonium and minor actinide transmutation in existing light water reactors and evaluates a variety of transmutation fuels to provide a common basis for comparison and to determine if any single target fuel provides superior transmutation properties. A model developed using the NEWT module in the SCALE 6.1 code package provided performance data for the burnup of the target fuel rods in the present study. The target fuels (MOX, PuO2, Pu3Si2, PuN, PuUZrH, PuZrH, PuZrHTh, and PuZrO2) are evaluated over a 1400 Effective Full Power Days (EFPD) interval to ensure each assembly remained critical over the entire burnup period. The MOX (5wt% PuO2), Pu0.31ZrH1.6Th1.08, and PuZrO2MgO (8wt% Pu) fuels result in the highest rate of plutonium transmutation with the lowest rate of curium-244 production. This study selected eleven different burnable absorbers (B4C, CdO, Dy2O3, Er2O3, Eu2O3, Gd2O3, HfO2, In2O3, Lu2O3, Sm2O3, and TaC) for evaluation as spectral shift absorber coatings on the outside of the fuel pellets to determine if an absorber coating can improve the transmutation properties of the target fuels. The PuZrO2MgO (8wt% Pu) target fuel with a coating of Lu2O3 resulted in the highest rate of plutonium transmutation with the greatest reduction in curium-244 production.

Application of the Pyrochemical DOS, Developed by the CEA, within Reprocessing of CERCER Transmutation Fuel Targets

Pyrochemical technology using high-temperature molten salts and molten metal media presents a potential interest for an overall separation and transmutation strategy for long-lived radionuclides. Within the frame of the two French acts on radioactive waste management, a pyrochemical R&D program was launched at the CEA Marcoule in the late 90's. The second step is the actinides back-extraction, which consists in a liquid/liquid oxidative stripping of the An from aluminium matrix into molten chloride media. The DOS process has been successfully demonstrated for treatment of oxide type fuels within the last years: the core of the process has been already assessed and the studies have shown high selectivity and a quantitative recovery of actinides.Within the framework of the SACSESS European research program, the pyrochemical activities focused on applications of the DOS process to reprocess CERCER transmutation targets. These particular type of fuels consist of a mixture of minor actinides (MA) oxides diluted in an inert (oxide MgO) matrices. The behaviour of matrices material was first investigated regarding the solubility in the fluoride salt, starting from both oxide powders or sintered pellets. Further investigations focused on the reductive extraction behaviour of Mg in metallic Al. The impact of Mg on the efficiency of An reductive extraction was also studied using uranium simulating the behaviour of MA oxides.

Pyrochemical treatment of spent nitride fuels for MA transmutation

Nitride fuels have several advantages including high thermal conductivity and high metal density (like metallic fuels) and high melting point and isotropic crystal structure (like oxide fuels). Since the late 1990s, the partitioning and transmutation of minor actinides (MA) has been studied to decrease the long-term radio-toxicity of high-level waste and to mitigate the burden of final disposal. Japan Atomic Energy Agency (JAEA) has proposed a dedicated transmutation cycle using an accelerator-driven system (ADS) with nitride fuels containing MA. The nitride fuel cycle we have developed includes a pyrochemical process. Our focus is on the electrolysis of nitride fuels and their refabrication from the recovered actinides; other processes are similar to the technology for metal fuel treatment and have been studied elsewhere. Here, we summarize our activity on the development of the pyrochemical treatment of spent nitride fuels.

XRD Monitoring of α Self-Irradiation in Uranium–Americium Mixed Oxides

The structural evolution under (241)Am self-irradiation of U(1-x)Am(x)O(2±δ) transmutation fuels (with x ≤ 0.5) was studied by X-ray diffraction (XRD). Samples first underwent a preliminary heat treatment performed under a reducing atmosphere (Ar/H2(4%)) aiming to recover the previously accumulated structural defects. Over all measurements (carried out over up to a full year and for integrated doses up to 1.5 × 10(18) α-decay events·g(-1)), only fluorite U(1-x)Am(x)O(2±δ) solid solutions were observed. Within a few days after the end of the heat treatment, each of the five studied samples was slowly oxidized as a consequence of their move to air atmosphere, which is evidenced by XRD by an initial sharp decrease of the unit cell parameter. For the compounds with x ≤ 0.15, this oxidation occurred without any phase transitions, but for U0.6Am0.4O(2±δ) and U0.5Am0.5O(2±δ), this process is accompanied by a transition from a first fluorite solid solution to a second oxidized one, as the latter is thermodynamically stable in ambient conditions. In the meantime and after the oxidation process, (241)Am α self-irradiation caused a structural swelling up to ∼0.8 vol %, independently of the sample composition. The kinetic constants of swelling were also determined by regression of experimental data and are, as expected, dependent on x and thus on the dose rate. The normalization of these kinetic constants by sample α-activity, however, leads to very close swelling rates among the samples. Finally, evolutions of microstrain and crystallite size were also monitored, but for the considered dose rates and cumulated doses, α self-irradiation was found, within the limits of the diffractometer used, to have almost no impact on these characteristics. Microstrain was found to be influenced instead by the americium content in the materials (i.e., by the impurities associated with americium starting material and the increase of cationic charge heterogeneity with increasing americium content).

Multirecycling of Pu, Am and Cm in BWR

We have investigated neutronics aspects of multirecycling of Pu, Am and Cm in BWRs, employing three uranium and three thorium-supported transmutation fuels. Our results show, that thorium-based cores allow for higher shares of MA in the fuel and thereby higher MA incineration without encountering a positive total void worth at any point of the multirecycling. In the uranium-based configuration the total void worth sets the limit on the MA share around 2.45%. The thorium-based fuels also exhibit a stronger Doppler feedback and somewhat degraded reactivity as compared to uranium-fuels. The alpha-heating in the fuel reaches equilibrium after six cycles, maintaining values of 24–31W/kgFUEL in the uranium-based configurations and 32–37W/kgFUEL in the thorium-based configurations. The neutron emission keeps rising through the multirecycling, the maximum value reached in the XV cycle ranges from 1.4×106 to 1.7×106n/s/g for uranium fuels and 2×106n/s/g for the thorium-based fuels.

Recent progress on minor-actinide-bearing oxide fuel fabrication at CEA Marcoule

Partitioning and transmutation (P&T) of minor actinides (MA: americium, neptunium and curium) in fast neutron reactors or accelerator-driven systems is a route envisaged to reduce nuclear waste inventory. Over the years, several modes of P&T were proposed, each being based on the use of dedicated fuels such as inert-matrix fuels, MA-bearing MOX or MA-bearing blankets. In this context, progress on the manufacturing of such fuels is a key-challenge in order to render P&T viable at the industrial scale. Here, MA-bearing oxide fuel fabrication and characterization conducted in the CEA Marcoule Atalante facility is reviewed. A particular attention is also given to the research conducted on uranium–americium mixed-oxides fuels, which are now considered the reference fuels for MA transmutation in France.

Moderating Material to Compensate the Drawback of High Minor Actinide Containing Transmutation Fuel on the Feedback Effects in SFR Cores

The use of fine distributed moderating material to enhance the feedback effects and to reduce the sodium void effecting SFRs is described. The drawback on the feedback effects due to the introduction of minor actinides into SFR fuel is analyzed. The possibility of compensation of the effect of the minor actinides on the feedback effects by the use of fine distributed moderating material is demonstrated. The consequences of the introduction of fine distributed moderating material into fuel assemblies with fuel configurations foreseen for minor actinide transmutation are analyzed, and the positive effects on the transmutation efficiency are shown. Finally, the possible increase of the Americium content to improve the transmutation efficiency is discussed, the limit value of Americium is determined, and the possibilities given by an increase of the hydrogen content are analyzed.

In Situ Study of the Solid-State Formation of U1–xAmxO2±δ Solid Solution

In order to reduce the nuclear waste inventory and radiotoxicity, U(1-x)Am(x)O(2±δ) materials are promising fuels for heterogeneous transmutation. In this context, they are generally fabricated from UO(2+δ) and AmO(2-δ) dioxide powders. In the subsequent solid solution, americium is assumed to be trivalent whereas uranium exhibits a mixed-valence (+IV/+V) state. However, no formation mechanisms were ever evidenced and, more particularly, it was not possible to know whether the reduction of Am(IV) to Am(III) occurs before the solid-solution formation, or only once it is established. In this study, we used high-temperature X-ray diffraction on a UO(2±δ)/AmO(2-δ) (15 mol %) mixture to observe in situ the formation of the U(1-x)Am(x)O(2±δ) solid solution. We show that UO(2+δ) is, at relatively low temperature (<700 K), oxidized to U(4)O(9-δ), which is likely to be caused by oxygen release from the simultaneous AmO(2-δ) reduction to cubic Am(2)O(3±δ). Cubic Am(2)O(3+δ) then transforms to hexagonal Am(2)O(3) at 1300 K. Thus, the initial Am(IV) is fully reduced to Am(III) before the solid solution starts forming at 1740 K. The UO(2) fluorite phase vanishes after 4 h at 1970 K, indicating that the formation of the solid solution is completed, which proves that this solid solution is formed after the complete reduction of Am(IV) to Am(III).

ECRIX-H experiment: Synthesis of post-irradiation examinations and simulations

The purpose of the ECRIX-H experiment is to study the behaviour of a composite ceramic target made of AmO 1.62 microdispersed in an MgO matrix irradiated for 318 EFPD in the Phenix sodium-cooled fast reactor (SFR), in a specific carrier sub-assembly equipped with annular blocks of CaH x acting as a neutron moderator. Results indicate that magnesia-based inert matrix targets display satisfactory behaviour and moderate swelling under irradiation, even for significant quantities of helium produced and a high burn-up. On this basis, the design of transmutation fuel pins for recycling of minor actinides (MA) in accelerator-driven systems (ADS) or in fast neutron reactors (FR) could be optimised so as to increase their performance level (initial MA content, burn-up, etc.). The measured Am fission rate (25 at.%) was found to be lower than that predicted by neutronic simulations probably due to the inaccuracies linked to the complexity of neutron modelling and the uncertainties on nuclear data related to moderated neutron spectrum. In addition, as most of the initial Am transmuted into Pu under irradiation, a PuO x -type phase was created within the initial AmO 1.62 particles, leading to the incomplete dissolution of the irradiated targets under standard reprocessing conditions. This issue will have to be considered and investigated in greater detail for all transmutation fuels and targets devoted to the multi-recycling of MA.

Minor Actinide Bearing Fuels: Fabrication and Irradiation Experience in Europe

Minor actinides constitute a significant contribution to the long term radiotoxicity of spent fuel. Their separation from spent fuel, fabrication and irradiation in homogeneous or heterogeneous reactor recycling modes provides a means to dramatically reduce their impact. Several challenges must be overcome in the separation and fabrication steps. Advanced robust fabrication processes are required to minimise dust and to produce fuels specifically designed to optimise their destruction through neutron irradiation. At the Institute for Transuranium Elements fabrication routes based on sol gel and infiltration have been developed to meet these demands. Dedicated fuels and targets have been produced and their irradiation behaviour is being tested. Fuels based on MOX and UO2 are being tested with respect to Gen IV fast reactor initiatives. Furthermore, dedicated minor actinide transmutation fuels based on the inert matrix concept (e.g. (Zr,Y,Am)O2 and Mo based (Pu,Am)O2 composite fuels) have been produced for application in the double strata concept, whereby the minor actinides should be transmuted in an ADS. Achievements and current status of these programmesare presented and discussed in this paper.

Neutronic Evaluation of a MHR System to Transmutation of Minor Actinides

The goal is to simulate the modular helium reactor (MHR) core to analyze the neutronic parameters behavior due the insertion of Pu isotopes and minor actinides (MAs) using shuffling scheme without compromising the safety parameters. Initially the core is filled with driver fuel (DF). After the burn-up, these fuels are then reprocessed and used to produce the transmutation fuel (TF). Some cycles after, the core is filled with DF and TF fuels. DF fuel is composed of Pu and Np while TF fuel is a mixture of Pu and MAs. The shuffling scheme was evaluated after each cycle. It was verified that neutronic parameters and isotopic composition reach equilibrium when this scheme is used. The WIMS code was used to perform the simulations and the following neutronic parameters were evaluated: infinite multiplication factor, spectrum hardening, and reactivity temperature coefficients.

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% 235U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Abstract Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R&D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic–ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic–metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg HM discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW (th) loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life. Presented results were used for testing newly-developed models installed in the TRANSURANUS code to deal with such innovative fuels and T91 steel cladding. Agreement among codes predictions was satisfactory for fuel and cladding temperatures, pellet-cladding gap and mechanical stresses.