Pure UO2 Research Articles

Transient behavior of oxide fuels with controlled microstructure and Cr2O3 additive

Microstructure and Cr2O3 doping profoundly impact the thermal-mechanical properties and fracture of oxides fuels. It is a challenge to study the transient behavior of nuclear fuels under loss-of-coolant-event (LOCA). In this study, the crack behavior of UO2 pellets with controlled grain structure and Cr2O3 doping was tested with rapid power ramping (300−900 °C per min) mimicking a prototypical LOCA heating profile. Dense micron-sized UO2 pellets display well-maintained integrity without cracking with the ramping up to 1500 °C at a heating rate of 8 °C per second. Fracture occurs in both pure and Cr2O3-doped dense nano-sized UO2 pellets. The Cr2O3 doped oxide fuel pellet with a larger grain size (~ 22.2 μm) displays the best performance under LOCA testing due to its highest thermal conductivity under high temperature. FEA calculations suggest a temperature gradient across the fuel pellet during transient testing, resulting in residual stress and cracking, which can be correlated with their thermal-mechanical properties.

A novel class of ATF fuels with large grain size, enhanced thermophysical properties and oxidation resistance

Two strategies have been extensively employed to develop advanced accident tolerant fuels (ATF): improving thermal conductivity and producing large grain sized pellets. However, there are few reports on the simultaneous utilization of both strategies. In this work, we fabricated Mo–Cr alloy reinforced UO2 (UMC) composite pellet with both high thermal conductivity and large grain size by a simple in-situ alloying method for the first time. The average grain size of UO2 increased from 6 μm to 113 μm. Thermal conductivity of the UMC pellet (with 2 vol% Mo) at 1200 °C increased by 36.46% compared with pure UO2. The increase rate of thermal conductivity per 1 vol% dopant reached about 18%. The Mo–Cr alloy formed continuous micro-cell layer around UO2 particles. Such continuous micro-cell Mo–Cr alloy layer and increase of UO2 grain size both helped improve the thermal conductivity of UMC composite pellets. Thermal expansion coefficient of the UMC under the operating temperature decreased by 11.90% compared with pure UO2. The oxidation resistance of the UMC pellet under high temperature steam was also improved. This work provides a new strategy towards fabricating ATF with both high thermal conductivity and large grain size.

Combined experimental and theoretical studies of some uranium(VI) complexes derived from imidazole-based carbenes

A series of bidentate, viz., 1,1'-(1,2-ethylene)-3,3'-dimethyldiimidazoline-2,2'-diylidene (L1), 1-methyl-3-(2-pyridylmethyl)-imidazoline-2-ylidene (L2) and tridentate, viz., 1,3-bis(2-pyridyl)-imidazoline-2-ylidene (L3) ligands have been obtained from their corresponding imidazolium salts through deprotonation reactions. Treatment of UO2Cl2(THF)3 with one equivalent L3 produces air-stable U(VI)-carbene complex 3, characterized by elemental analysis, FTIR, NMR spectroscopy as well as extended X-ray absorption fine structure (EXAFS) analysis. EXAFS result indicates that the ligand is bonded through one C and two N-atoms to the uranium atom. Attempts to synthesize uranyl complexes derived from L1 and L2 were also successful but these complexes are decomposed quickly within one hour at room temperature. 3 produces pure UO2 powder when heated under an argon atmosphere from room temperature to 600 °C with constant heating rate of 5 °C/min. The solid-state UV–Vis spectrum of the compound shows absorption peaks at 332 and 450 nm. The excitation spectrum of 3 (at λ em = 520 nm) exhibits two almost symmetrical peaks at 273 and 368 nm. Density functional theory-based quantum mechanical calculations indicate that a partial covalent interaction exists between the carbene C and U while a weak non-covalent interaction exists between carbene N and U atoms.

The effect of cerium, neodymium, and ytterbium doping on UO2 dissolution

The dissolution rate of spent nuclear fuel under environmental conditions has been studied heavily to understand the impacts of a failed waste package on potential radionuclide release from a geologic repository. Multiple countries have evaluated these scenarios in oxidizing and reducing environments respective to their relevant repository conditions. The repository environment (e.g., water chemistry, temperature, oxygen content) and the fuel itself (i.e., chemical composition of the fuel) both heavily impact the dissolution rate of the spent fuel. This work examined the impact that rare earth element dopants (Ce, Nd, Yb) have on the fuel dissolution under repository relevant conditions with decreasing oxidizing conditions using a single pass flowthrough system. UO2 samples were doped with Ce, Nd, or Yb with concentrations between 1 and 5 at%. The addition of dopants to the samples reduced the dissolution rates on most samples relative to pure UO2 samples. Preliminary experiments that occurred in a less oxidizing environment showed a reduction in dissolution rate compared to fully oxidizing conditions. The results within highlight the importance of dopant behavior in used fuel dissolution modeling.

Study on the electrodeposition of uranium in chloride molten salt.

This study focuses on the recovery of UO2 from oxide spent fuel using electrodeposition. U3O8 was used as the initial material and dissolved in NaCl-2CsCl using NH4Cl at high temperatures by means of chlorination reaction. The electrolysis process was conducted using a three-electrode system to investigate the effects of cathode material and diameter, electrolysis temperature, electrolysis time, electrolysis voltage, and uranium concentration in the molten salt on the electrolysis reaction. By optimizing the electrolysis conditions, pure UO2 with a recovery efficiency of 97% was obtained, and the products were characterized using XRD, SEM-EDS, ICP-AES and XPS. It was found that within the scope of this experiment, increasing the cathode diameter, extending the electrolysis time, and increasing the reduction voltage appropriately all led to an improvement in the recovery efficiency of the electrolysis reaction, while other conditions had minimal effect on the reaction. Furthermore, doping of the electrolyte system was performed by adding La, Ce and Nd elements, while the removal of La showed good purification effects, with a maximum decontamination factor of 119. Furthermore, the system showed good purification effects for Nd, with a decontamination factor of 57.

Computational study of thermophysical properties of cerium doped UO2: Effect of oxidation states

The interaction of fission products with UO2 fuel, which is generated during the nuclear fission, can amend the fuel behavior. Among different lanthanide (Ln) fission products produced, Cerium (Ce) shows considerable yield. The Ce substitution in UO2 with isovalent Ce4+ cation is interesting due to the absence of oxygen vacancies and no change in the oxidation state of U atoms in the doped system. In this study, the trivalent and tetravalent Ce atom substitution in UO2 and their effect on thermophysical properties is investigated using density functional theory. Further, the effect of Ce doping concentrations in UO2 is assessed by varying the Ce ion concentrations in the lattice (6.25 %, 12.5 %, 25 %, and 50 %) for various charge balancing mechanisms. The lattice volume of Ce doped UO2 structures show higher or lower values compared to undoped UO2 depending upon the oxidation state of Ce and U atoms. The bulk modulus tends to reduce for Ce substituted UO2 as compared to pure UO2. The electronic density of states analysis shows that Ce substitution levels as well as the oxidation state of Ce/U atoms strongly influences the band structure of the UO2 lattice. The specific heat capacity of Ce doped UO2 structures is calculated at different Ce doping concentrations and for disparate charge balancing procedures using the quasi harmonic approximation(QHA). Present results show good agreement with the experimental values reported for similar Ce substituted UO2 structures. Thus, the systematic investigation with varying Ce fission product concentrations and oxidation states of cations (Ce and U) can give insight into the fuel properties of UO2 under reactor conditions.

Thermodynamic model of the oxidation of Ln-doped UO2

Ln-doped UO2 is often considered as a model system of spent nuclear fuel (SNF) helping to reveal effects of fission and activation products on its chemical stability. Comparing thermodynamics of UO2-UO3 and LnO1.5-UO2-UO3 systems provides a means to understand the phenomenon of an increased resistivity of Ln-doped UO2 to oxidation in air relative to pure UO2. Here a thermodynamic model is developed and is applied to investigate detailed phase changes occurring along the oxidation of Ln-doped fluorite to U3O8. The study proposes that an enhanced resistivity to oxidation of Ln-doped UO2 is likely caused by a thermodynamically driven partitioning of Ln between a fluorite-type phase and a U3O8 polymorph, which at ambient temperatures becomes hindered by slow diffusion.

Densification kinetics and sintering behavior of UO2 and 0.5 wt.%MnO‐doped UO2

AbstractIn this work, the sintering kinetics of pure UO2 and 0.5 wt.%MnO‐doped UO2 was studied by a high‐temperature dilatometer heated up to 1500°C. In addition, the sintering behavior of pure UO2 and 0.5 wt.%MnO‐doped UO2 was studied by pressureless sintering technique. The results showed that MnO doping enhanced the grain boundary diffusion of UO2, which can effectively decrease the densification temperature and promote grain growth. The sintering temperature of UO2 was significantly reduced by about 200°C with the addition of 0.5 wt.%MnO. The microscopic morphology studies showed that there were still fine particles agglomerated, forming sintered spheres in the matrix even if no severe agglomeration and bimodal size distribution were observed in raw UO2 powder. The microstructure evolution of the sintered sphere and UO2 matrix during the densification process was studied by isothermal sintering. Finally, the present analyses indicated that the densification of UO2 matrix can be accelerated by adding MnO or increasing the sintering temperature, thus improving the densification inhomogeneity of UO2 matrix.

Thermal conductivity of UO2 pellets enhanced by a semi-continuous structure of Ti3SiC2

The current distribution structure of the thermal conductivity (TC) enhancements in UO2 matrix can be divided into three categories: dispersed structure, continuous structure and oriented structure. This paper develops a new semi-continuous structure to improve the TC of UO2. Ti3SiC2 is selected as the enhancement in this paper due to its excellent irradiation resistance and high TC at elevated temperature. The semi-continuous structure of Ti3SiC2 is formed through a simple centrifugal spheroidization of UO2 and a followed powder mixing process. Both the theoretical prediction and experimental results indicate the semi-continuous structure has a close TC enhancing efficiency to the continuous structure. Meanwhile, the semi-continuous structure has the advantages of simplifying the fabrication process and can be applied to more kinds of enhancements compared to the continuous structure. The TC of Ti3SiC2/UO2 is remarkably improved at elevated temperature compared with pure UO2, which is beneficial to the fuel safety.

Accelerated dissolution of doped UO2-based model systems as analogues for modern spent nuclear fuel under repository conditions

Systematic single-effect dissolution studies were carried out on the dissolution behaviour of pure and Cr- or Nd-doped UO2 reference pellets as model materials for spent nuclear fuel with varying doping levels, densities, and grain sizes as well as of industrially produced Cr- and Cr/Nd-doped UO2 pellets. The results were obtained from accelerated static batch dissolution experiments performed under strictly controlled conditions using H2O2 as simulant for radiolytic oxidants formed due to the alpha-irradiation of water. The results indicate that the addition of Cr and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behaviour of these model materials compared to pure UO2 reference materials. Contrarily, the dissolution rates of Nd-doped pellets are significantly lower than those of pure and Cr-doped pellets. These results provide additional insights into the influence of doping on the dissolution behaviour of modern spent light water reactor fuels under the post-closure conditions expected in a deep geological repository.Graphical abstract

Exploring a surrogate of Pellet–Cladding interaction: Characterization and oxidation behavior

The present work analyzes the effect of different Zr contents on the microstructure and thermal behavior of non-irradiated UO2. A set of Zr-doped UO2 (0, 20, 40, 80, and 100 wt%) pellets was prepared via solid-state synthesis by mimicking the chemical bonding between ZrO2 and UO2. After sintering, the Zr-doped UO2 monoliths were characterized by scanning electron microscopy, Raman spectroscopy, and X-ray diffraction. Oxidation of the Zr-doped UO2 samples has been monitored by thermogravimetric analysis. Results of thermal behavior under air and low oxygen partial pressure show a profound effect of delayed oxidation with the addition of Zr to UO2 and then an increased oxidation resistance of UO2 to U3O8 when compared to pure UO2 pellet.Graphical abstract

Simulation of the Melting Behavior of the UO2-Zircaloy Fuel Cladding System by Laser Heating

The current research focuses on laser melting and successive analysis of laboratory-scale uranium dioxide nuclear fuel samples in direct contact with Zircaloy-4 cladding. The goal was to characterize the melted and refrozen interfaces, in particular, observing local changes of the melting point and interdiffusion of fuel and cladding materials under inert gas (Ar), in the presence of hydrogen (Ar + 6% H2) or in air. Results obtained by laser heating UO2 pellets clad in a Zircaloy ring were interpreted in light of reference tests performed on pellets in which UO2 and zirconium were blended in a series of given compositions. The sample composition was analyzed by scanning electron microscopy to verify the occurrence of diffusion and segregation phenomena during the laser-heating cycles. Laser-melting experiments were performed on pellets of uranium dioxide clad in Zircaloy-4 rings to simulate the configuration of a light water reactor fuel rod. Under inert gas, the material interdiffusion resulted in consistent melting point depression (of up to 200 K below the melting point of pure UO2) at the interface between the fuel and the cladding. Experiments carried out in the presence of H2 displayed a more limited effect on the melting temperature, but they resulted in a remarkable embrittlement of the whole structure, with large fragmentation of the Zircaloy cladding. This was probably due to the formation of brittle and highly volatile Zr hydrides. The observed melting point decrease was even more pronounced (up to over 400 K) under air in uranium-rich samples, due to the change in the stoichiometry of UO2 in UO2+x.

Simulated advanced gas-cooled reactor spent nuclear fuels: Determination of the O/U ratio - an XRD, XPS and Raman study

This paper reports a detailed chemical and materials characterization study of novel Simulated Spent Nuclear Fuels (SIMFuel) which replicate the chemical and microstructural state of Spent Nuclear Fuel (SNF) discharged from a UK Advanced Gas-cooled Reactor (AGR). Recent advances in the analysis of pure UO2 samples by XPS, micro-Raman spectroscopy and XRD, have been assessed for deployment in the characterisation of SIMFuels generally and these AGR SIMFuels in particular. All three analytical methods reveal the extent that the UO2 bulk matrix is defected in the SIMFuels by the presence of lanthanide dopants. XPS, by inspection of the U4f peaks and their satellites, indicates the presence of U(V) in the UO2 matrix as a means to charge compensate for the incorporation of Ln(III) states and presence of a slight hyperstoichiometry in the UO2 solid solution. This was corroborated by detailed Raman analysis of the UO2 matrix of the SIMFuels – wherein the (Ba,Sr)ZrO3 and metallic particle phases, that simulate the precipitated grey phases and ε-particles formed in real spent fuel, were avoided – which indicated the presence of both interstitial oxygens and oxygen vacancies, the latter as a parallel means of charge compensation for the presence of Ln(III). Further confirmation was provided by XRD measurements through observation of a lattice parameter contraction arising from the presence of the smaller U(V) ion in the U(IV)O2 matrix. XPS and Raman allow for the reporting of O/U ratios in the SIMFuels. These XPS and Raman-derived O/U ratios are in good agreement, indicating: (i) that the SIMFuels are near stoichiometric/slightly hyperstoichiometric, concurring with previous analogous studies on real and simulated light water reactor fuels; and (ii) that either technique may find application in the analysis of real fuel analysis, both pre- and post-irradiation.

Impact of ruthenium metallic particles on the dissolution of UO2 in nitric acid

UO2 pellets incorporating 3 mol.% of Ru was prepared by using a wet chemistry route and then characterised. The speciation, morphology, as well as spatial distribution of Ru in the sintered samples, were determined. The synthesised samples were submitted to dissolution tests in 0.1 M nitric acid at 60 °C and the dissolution of pure UO2 pellets was also studied with and without the presence of Ru metallic particles in the solution. The evolution of the U, Ru, and nitrous acid concentrations in solution was measured and the residues of dissolution were further characterised. The obtained results unambiguously demonstrated the catalytic activity of Ru-metal particles during UO2 dissolution in nitric acid provided that a solid/solid interface existed between UO2 and Ru-metal particles. This positive impact on the dissolution kinetics of UO2 was supported by redox reactions taking place at both nitric acid solution/Ru-metal particles and at Ru-metal particles/UO2 interfaces.

Structure, mechanical and thermo-physical properties of lanthanide fission product doped UO2 in U(V) state: A density functional study

During the nuclear fission of UO2 fuel, many fission products are generated which are trapped in the fuel matrix. These fission products can amend the mechanical and thermo-physical properties of UO2 fuel, which are a prime concern for nuclear safety. In this regard, we investigate the effect of different lanthanide (Ln) fission products on structural, mechanical, electronic and thermal properties of UO2. The substitution of U4+ ions by Ln3+ ions creates a charge imbalance in the lattice which will be neutralized by either oxidation of U atoms from +4 to +5 state (U2Ln2O8) or creation of oxygen vacancies (U2Ln2O7). Here, we study the former case, where the U(IV) ions are oxidized to U(V) state in the Ln substituted UO2 (Ln- UO2) fuel using density functional theory with Hubbard-U corrections. The volume of different Ln-UO2 structures is calculated and compared with undoped UO2. The mechanical properties such as bulk modulus and Young's modulus are calculated from the optimized structures of Ln-doped UO2 and show deviation from undoped UO2 and Ln-doped UO2 with oxygen vacancies. The total density of states and projected density of states indicate that the presence of Ln fission products alters the electronic properties of UO2. The thermal property such as specific heat capacity is evaluated at different temperatures using quasi-harmonic approximation applied to phonon frequencies obtained from density functional perturbation theory. We do observe an excellent agreement of the calculated specific heat capacity with the experimental values for the pure UO2. The counter balancing effect of heat capacity values of U2Ln2O8 and U2Ln2O7 systems may control the net heat capacity of the UO2 fuel matrix.

Hydrothermal synthesis of UO2 nanoparticles and their formation mechanism

We report a simple and general method for the production of UO2 nanoparticles by heat-treating an aqueous mixture of uranyl acetate dihydrate and ethylenediamine. The products were characterized, and the effects of heat treatment temperature and time were investigated. When heat treatment time was fixed at 24 h, UO2 could not be obtained at temperature lower than 160 °C, both UO2 and ianthinite were obtained at 200 and 240 °C, while pure UO2 nanoparticles were collected at 280 °C and higher temperatures. The average size of these nanoparticles varied in the range of 250–450 nm, which could be adjusted by tuning the reaction temperature and incubation time. Longer incubation time led to slight size increase due to Ostwald ripening. The mixture precursor took shorter time to transform to UO2 at higher temperatures. A possible synthesis mechanism was proposed, including three stages: (I) initial precursor formation and transformation, (II) mid-term formation of ianthinite and UO2, and (III) the final formation of UO2 nanoparticles. This research provides a facile method for the synthesis of UO2 nanoparticles with excellent size stability in a relatively wide range of experimental conditions. Moreover, this method provides a new way for synthesis of ianthinite, a rare and hard-to-synthesize hydrous uranyl oxide.

Reassessment of Computational Tools for the Modeling of Heat Transfer in a Molten UO2 Pool in Natural Convection

Abstract This work, performed within the European sodium fast reactor safety measures assessment and research tools (ESFR-SMART) H2020 European project, is part of a larger framework intending to reassess the modeling of heat transfer in molten pools on SCARABEE available experimental results. This paper presents simulation results of the in-pile Bain Fondu (BF1) test, performed within the SCARABEE-N program, using accident source term evaluation code (ASTEC), Sn method implicit multifiel multiphase Eulerien recriticality (SIMMER) III, and SIMMER V simulation tools as well as comparison with its available experimental data. This program was performed in the 1980s in the frame of the Safety Assessment studies of Superphenix sodium-cooled reactor. This test was dedicated to verify the stability of a pure molten UO2 pool under decay heat conditions within natural convection and the long-term resilience of the peripheral fuel crust. The pool was generated in a stainless steel crucible by a progressive heating of a fuel pellet stack through six successive power plateaus. For the benchmark purposes, only the molten pool behavior k the last power plateau, where the pool was the largest and the fuel temperatures the highest, was investigated. Experimental data such as the axial profile of radial heat fluxes and heat transfer from the pool to the surrounding interassembly coolant as well as the axial profile of the peripheral fuel crust thickness were used for the reassessment of the simulation tools. In addition, other variables of interest not measured during the test, such as the radial and axial velocities in the pool, were also benchmarked. Finally, a critical analysis of the correlations and models used in the different simulation tools for the BF1 test modeling is also provided in the paper.

Evaluation of the corrosion behavior of modern spent nuclear fuels under repository conditions

Abstract. In Germany it is planned to directly dispose spent nuclear fuel (SNF) from nuclear power plants together with other high-level radioactive wastes (HLW) from former SNF reprocessing (e.g., vitrified waste), in a deep geological repository for heat-generating wastes – the siting process for this repository was started in 2017 and is ongoing. Based on several decades of research, development, and demonstration (RD&D) it is generally accepted at the technical and scientific level that direct disposal of HLW and SNF in deep mined geological repositories is the safest and most sustainable option (CEC, 2011; IAEA, 2004). The current efforts to improve the performance and accident tolerance of fuels in nuclear power generation resulted in an increased utilization of a variety of new types of light-water reactor (LWR) fuels such as fuels doped with Cr, Al, and Si. This doping leads to a significant change of the microstructure of the fuel matrix. The corrosion behavior of these types of fuels under conditions relevant to deep geological disposal has hardly been studied so far; however, this is of crucial importance as the development of a robust safety case for deep geological disposal of SNF requires a solid understanding of its dissolution behavior over very long time scales (up to 1 million years). To fill this knowledge gap, additional systematic studies on modern doped UO2 fuels were needed. Corrosion experiments with SNF cannot entirely unravel all of the various concurring effects of the dissolution mechanism due to the chemical and structural complexity of SNF and its high beta and gamma radiation field during the first 1000 years; moreover, technical restrictions only allow a very limited number of experiments. Therefore, within the EU-DisCo project (https://www.disco-h2020.eu, last access: 11 October 2021), a very ambitious programme of corrosion studies on irradiated Cr and Al/Cr doped fuels was carried out, which was complemented by systematic single-effect dissolution studies (e.g., with respect to doping level, grain size and thermodynamic aspects) performed on carefully prepared and characterized, simplified UO2-based model materials. Here, we present recent results on the dissolution behavior of tailor-made UO2 model materials in accelerated static batch experiments using H2O2 as simulant for radiolytic oxidants, present in long-term disposal scenarios for SNF in failed container conditions due to the alpha irradiation of water. In these dissolution experiments pure UO2 reference pellets exhibiting different densities and grain sizes, as well as Cr-doped UO2 pellets with various Cr-doping levels, produced using different doping methods having different grain sizes, were used. In addition, Nd-doped and industrially produced Cr- and Cr/Nd-doped UO2 pellets were used to determine the influence of these parameters on the dissolution rates. The dissolution experiments were performed under strictly controlled conditions with respect to exclusion of oxygen, temperature control, and exclusion of light. This bottom-up approach was followed to understand how the addition of Cr-oxide into the fuel matrix affects SNF dissolution behavior under repository relevant conditions. The results of the dissolution experiments performed with real SNF and the model materials obtained by the DisCo partners build the basis for numerical simulations on the dissolution behavior of modern SNF. First results of the data evaluation indicate that the addition of dopants and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behavior of these fuels under repository relevant conditions compared to standard SNF (i.e. dissolution rates agree within an order of magnitude).

Uranium corrosion characterization by handheld laser-induced breakdown spectroscopy

A commercial-off-the-shelf handheld laser-induced breakdown spectroscopy system was modified, calibrated and tested for discrimination between uranium oxide and uranium hydride when assessing uranium corrosion products. Operating parameters, system modifications, and onboard data analysis were developed to accomplish the analytical task. This work represents the first laser-induced breakdown spectroscopy analysis of bulk uranium hydride, a pyrophoric material, within an inert glovebox. Finally, the work shown here illustrates a rapid identification technique for uranium corrosion types from pure UO2 or U3O8 to pure UH3.

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY

STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY. NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core.Keywords: Criticality, safety parameters, NuScale fuel assembly, MCNP6, ENDF/B-VII.