Fully Ceramic Microencapsulated Research Articles

Numerical study of TRISO particles with random size and location distribution in cuboid and cylindrical matrix: A validation for two-regime heat conduction model

During reactor operation, the precise size values or specific size distribution of tri-structural isotropic (TRISO) fuel particles randomly dispersed within the fully ceramic microencapsulated (FCM) fuel elements are often unknown. To investigate the impact of the random distribution of positions and sizes of TRISO particles within a cuboid or cylindrical matrix on the temperature distribution of fuel elements and to establish a practical heat conduction model, TRISO fuel particles with randomly distributed positions, diameters following Gaussian or uniform distributions are generated using relevant algorithms. A cuboid and a cylindrical TRISO fuel element are modeled in ANSYS, and a three-dimensional temperature field distribution of the fuel element is obtained through numerical simulation. The probability density of the maximum temperature of the fuel element is calculated and fitted. Finally, the applicability of the two-regime heat conduction model over a broader range is validated based on the simulated data. The conclusions of this study provide reference information for predicting the temperature distribution of FCM fuel elements containing TRISO particles with random size and location distribution.

Pressureless sintering of large sized fully ceramic microencapsulated fuel pellets

Large-sized fully ceramic microencapsulated (FCM) nuclear fuel pellets were fabricated by a novel tape casting process in this study. The process involved producing a single layer of FCM fuel green tapes embracing TRISO particles by tape casting with silicon carbide ceramic as the matrix. The resulting tapes were stacked vertically and made into green bodies through cold isostatic pressing. Large-sized FCM fuel pellet with diameters over 86 mm was then successfully prepared by pressure-less sintering with AlN-Y2O3-Sc2O3−MgO as sintering additives. The FCM fuel pellets obtained through this method exhibited a dense structure, flat surface, and no cracks. Thus, the integrated formation of the fuel-free zone and fuel-filled zone could be realized by precisely controlling the fuel-free zone thickness. The TRISO particles were regularly and uniformly distributed in a triangular shape and remained structurally intact. By adjusting the spacing of the TRISO particles in the FCM fuel pellets to a range of 500 to 1500 μm, the volume fraction of fuel particles in the FCM fuels was controlled between 12.8 and 31.7 vol.%.

Fabrication of beryllium oxide based fully ceramic microencapsulated nuclear fuels with dispersed TRISO particles by pressureless sintering method

In this research paper, two types of tristructural isotropic (TRISO) particles were dispersed within beryllium oxide (BeO) matrix using the pressureless sintering method to create fully ceramic microencapsulated (FCM) nuclear fuel. The findings indicated that TRISO particles with an outermost layer comprising silicon carbide (SiC) can establish interfacial bonding with BeO matrix, while those with an outermost layer comprising pyrocarbon (OPyC) and BeO matrix exhibit a noticeable gap after the sintering process. Results from transmission electron microscopy (TEM) revealed the presence of an amorphous phase transition zone between SiC and BeO. This zone, according to molecular dynamics (MD) simulation results, is an amorphous phase composed of Be, Si, and O atoms. Overall, TRISO particles with an outermost layer of SiC are more suitable for sintering with BeO matrix than those with an outermost layer of OPyC.

Multi-objective optimization design of TRISO-based fully ceramic microencapsulated fuel

Fully Ceramic Microencapsulated (FCM) fuel is a new type of material in nuclear energy field. It has become a promising accident tolerant fuel (ATF) candidate due to its advantages of high radiation stability, high fission product containment capacity and excellent thermo-mechanical performance. During operation, the temperature, stress, deformation and other key parameters will change significantly, which directly affect the safety of nuclear reactor. Therefore, it is necessary to carry out a multi-physics fully-coupled analysis suitable for the multi-scale, multi-component, multi-phase and nonlinear coupling processes in the reactor, so as to sort out the influences of material properties, radiation behavior, thermo-mechanical performance from numerous influencing factors. In this paper, based on the irradiation-thermo-mechanical coupling method, the performance of TRISO-based FCM fuel under irradiation conditions was analyzed, key thermo-mechanical parameters and other influencing factors of the fuel were obtained, failure behavior of the fuel element was evaluated. On this basis, the theoretical and numerical model of multi-objective optimization method were established, the optimum parameters including material properties, fuel geometry and particle arrangement were found by optimization method. The optimization results showed that the multi-objective optimization results could provide multiple Pareto optimal solutions for the FCM fuel. Specifically, when the TRISO particle spacing is 200 µm and 150 µm, the objective function difference between the optimal heat transfer solutions is only about 1.7%, meanwhile the mechanical properties vary significantly (the difference between the objective functions of the optimal solutions is about 12.4% ∼26.5%), indicating the response of stress is more sensitive than temperature. When the TRISO particles are tightly packed (particle spacing less than 100 µm), the difference of objective function between the optimal solutions is less than 5%, illustrating the optimization has limited improvement on the thermo-mechanical performance of the fuel, namely the optimization space is limited.

Fully ceramic microencapsulated fuels with high TRISO particles loading capacity fabricated by gel-casting

A novel gel-casting process for fabrication of the fully ceramic microencapsulated (FCM) fuel with high TRISO particles loading capacity was proposed. In the gel-casting process, silicon carbide (SiC) ceramic was used as matrix, and low toxic N-methylol acrylamide was used as organic monomer with appropriate proportion of cross-linking agent, dispersant, and initiator to achieve the strength of green bodies. FCM fuel pellets with high TRISO particles loading capacity were successfully prepared by SPS process. By using the gel-casting method, the volume fraction of TRISO particles could reach 57%, and the average distance between TRISO particles in a plane was 52.4 μm, which could avoid TRISO particles damage introduced by direct contact between the particles. The structure of SiC matrix was dense and complete, and there was a coherent contact between the matrix and TRISO particles, which ensured the performance and reliability of FCM fuel.

Multi-scale heat conduction models with improved equivalent thermal conductivity of TRISO fuel particles for FCM fuel

Fully Ceramic Microencapsulated (FCM) fuel is emerging advanced fuel material for the future nuclear reactors. The fuel pellet in the FCM fuel is composed of matrix and a large number of TRistructural-ISOtopic (TRISO) fuel particles which are randomly dispersed in the SiC matrix. The minimum layer thickness in a TRISO fuel particle is on the order of 10−5 m, and the length of the FCM pellet is on the order of 10−2 m. Hence, the heat transfer in the FCM pellet is a multi-scale phenomenon. In this study, three multi-scale heat conduction models including the Multi-region Layered (ML) model, Multi-region Non-layered (MN) model and Homogeneous model for FCM pellet were constructed. In the ML model, the random distributed TRISO fuel particles and coating layers are completely built. While the TRISO fuel particles with coating layers are homogenized in the MN model and the whole fuel pellet is taken as the homogenous material in the Homogeneous model.Taking the results by the ML model as the benchmark, the abilities of the MN model and Homogenous model to predict the maximum and average temperature were discussed. It was found that the MN model and the Homogenous model greatly underestimate the temperature of TRISO fuel particles. The reason is mainly that the conventional equivalent thermal conductivity (ETC) models do not take the internal heat source into account and are not suitable for the TRISO fuel particle. Then the improved ETCs considering internal heat source were derived. With the improved ETCs, the MN model is able to capture the peak temperature as well as the average temperature at a wide range of the linear powers (165 W/cm∼ 415 W/cm) and the packing fractions (20%–50%). With the improved ETCs, the Homogenous model is better to predict the average temperature at different linear powers and packing fractions, and able to predict the peak temperature at high packing fractions (45%–50%).

Effects of starting powder on microstructure and thermal conductivity of pressureless-sintered fully ceramic microencapsulated fuels

The effects of the starting SiC powder (α or β) with the addition of 5.67 wt% AlN–Y2O3–CeO2–MgO additives on the residual porosity and thermal conductivity of fully ceramic microencapsulated (FCM) fuels were investigated. FCM fuels containing ∼41 vol% and ∼37 vol% tristructural isotropic (TRISO) particles could be sintered at 1870 °C using α-SiC and β-SiC powders, respectively, via a pressureless sintering route. The residual porosities of the SiC matrices in the FCM fuels prepared using the α-SiC and β-SiC powders were 1.1% and 2.3%, respectively. The thermal conductivities of FCM pellets with ∼41 vol% and ∼37 vol% TRISO particles (prepared using the α-SiC and β-SiC powders, respectively) were 59 and 41 Wm−1K−1, respectively. The lower porosity and higher thermal conductivity of FCM fuels prepared using the α-SiC powder were attributed to the higher sinterability of the α-SiC powder than that of the β-SiC powder.

Heat transfer performance of the solid microencapsulated fuel in light water reactors

Solid Microencapsulated Fuel (SMF) is a kind of new accident-tolerant fuel (ATF) concept that disperses TRISO coated fuel particles directly into a type of matrix in the form of a solid rod. This paper aims to analyze the heat transfer performance of SMF based on a single pin, compared to the traditional fuel design in the 200 MW Nuclear Heating Reactor (NHR200-II). Two kinds of SMFs have been evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. The centerline temperature of the SMF rod at the beginning of life (BOL) was preliminarily calculated based on the effective thermal conductivity (ETC). The ETC was calculated by Maxwell, Chiew&Glandt, and d-EMT models, compared to the finite element models. The ETC of TRISO particle was also calculated before calculating that of the SMF rod because of its multilayer structure. To get a more realistic temperature distribution of SMF rod, a neutronic analysis was done by OpenMC by acquiring the power distribution of randomly distributed TRISO particles. Finally, a practical temperature distribution of SMF rod was achieved by the finite element model considering the random distribution of TRISO particles. This study showed that SMF has a better thermal transfer capacity than conventional fuel design and even fully ceramic microencapsulated (FCM) fuel because of its integral structure and higher matrix thermal conductivity. The helium gap in conventional fuel and FCM fuel would bring a temperature rise of nearly 150˚C in this case. The centerline temperature of SMF is much lower than the traditional fuel. A more refined temperature distribution of SMF rod based on the finite element model showed a lower maximum temperature than the ETC calculated. It also showed a less uniform temperature distribution because of the heat generation of fuel particles and the possible concentration of randomly distributed fuel particles.

Neutronic analysis on TRU multi-recycling in PWR-based SMR core loaded with MOX (U-TRU) and FCM (TRU) fueled assemblies

An analysis on TRU multi-recycling in a light water-cooled SMR core loaded with MOX-FCM fuel assemblies employing MOX and FCM (Fully Ceramic Micro-encapsulated) fuel rods was performed for TRU transmutation. We used a two-step procedure composed of McCARD/MASTER where the assembly-wise two group homogenized neutron cross sections are generated using a Monte Carlo depletion code, while the core analysis was performed using a two-group nodal diffusion code. The reload core calculation based on three-batch refueling scheme was conducted using a TRU recycling scheme in which TRU in MOX fuel rods are multi-recycled but TRUs in FCM fuel ones are sent to the final repository after the discharge without recycling. An overall TRU mass balance analysis through the 5th to 20th cycle showed that 21.1% of external TRU feed can be consumed by utilizing the light water-cooled SMR core without considerable degradation of safety-related parameters.

Research of the spatial effect of PWR fuel pin with disperse particles

Fully ceramic microencapsulated (FCM) fuel is one of the most popular accident tolerant fuels (ATF) and it has a particular spatial structure, made of SiC matrix embedded with TRISO fuel particles. Fuel pin embedded with disperse burnable poison particles, has similar spatial structure. They are both double-heterogeneous and they have much unique neutronic characteristic. In this paper, efforts are made to analyze how the spatial structure of the disperse fuel particles and burnable poison particles affects their neutronic behaviors in detail, from pin-cell level, which is also referred as spatial effect. Many important neutronic parameters are discussed, including infinite multiplication factor (Kinf), four factors (fast fission factor ∊, resonance escape probability factor p, thermal utilization factor f and thermal reproduction factor η), neutron flux level, nuclides number density. They are compared with an equivalent homogeneous structure and the corresponding differences would reveal the unique spatial effect of this kind of double-heterogeneous structure. Besides, sub-regions analyses are also carried on. Finally, it is proved that the double-heterogeneous spatial structure would bring much influence on neutronic behaviors, and the spatial effect of disperse burnable poison particles would be even more serious.

High volume packing fraction TRISO-based fuel in light water reactors

For a decade, fully ceramic microencapsulated (FCM) fuel, containing tri-structural isotropic (TRISO) fuel particles in a silicon carbide (SiC) matrix, has been investigated as an accident-tolerant fuel for light water reactors (LWRs). Other examples exist of TRISO-based concepts for LWR fuels with different matrix materials. Previous studies assumed TRISO particle volume packing of approximately 0.44 in SiC (or another) matrix, the highest realistic packing fractions possible with conventional manufacturing. Recent advances in advanced manufacturing have yielded the development and demonstration of a fuel form that consists of conventionally manufactured TRISO particles in a 3D-printed SiC matrix with significantly higher possible TRISO packing fractions (0.5–0.7). This increased uranium loading enhances the viability of using TRISO-based particle fuel forms in LWRs. The viability of high-packing-fraction TRISO-based particle fuel forms in LWRs is assessed from the perspective of fuel cycle length, achievable fuel burnup, reactivity coefficients, and fuel cycle performance. Higher-packing-fraction TRISO-based fuel enables either longer cycle lengths (by ∼25% at a packing fraction of 0.55 relative to 0.44) at a constant enrichment or decreased enrichments (by ∼25% at a packing fraction of 0.55 relative to 0.44) at a constant cycle length. Studies of different fuel kernel types (uranium nitride, uranium oxycarbide, and uranium carbide) yield similar results, although the cycle length of uranium oxycarbide is shorter than for uranium nitride or uranium carbide (due to the lower density of uranium oxide). This work also characterized the production of 14C resulting from neutron absorption in 14N during operation for uranium mononitride fuel kernels; the ratio of 14C/N was 1–2 at. % at discharge. For the fuel cycle evaluation, the activity of spent nuclear fuel and high-level waste at 100 and 100,000 years was lower for high-packing-fraction fuels than for conventional LWR fuel. Environmental impact metrics were similar overall, but higher on the front end of the fuel cycle and lower on the back end of the fuel cycle. Reactivity coefficients of higher-packing-fraction TRISO-based fuel were reasonable compared with those of conventional fuels.

The CANDLE burnup strategy applied to small modular pressurized water reactor loading with fully ceramic microencapsulated fuel

Abstract For post-Fukushima nuclear power plants, there has been interested in accident-tolerant fuel (ATF) since it has better tolerant in the event of a severe accident. The fully ceramic microencapsulated (FCM) fuel is one kind of the ATF materials. In this study, the small modular pressurized water reactor (PWR) loading with FCM fuels was investigated, and the modified Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor (CANDLE) burnup strategy was successfully applied to such compact reactor core. To obtain ideal CANDLE shape, it’s necessary to set the infinity or enough length of the core height, but that is impossible for small compact core setting infinity or enough length of the core height. Due to the compact and finite core, the equilibrium state can only be maintained short periods and is not obvious, other than infinitely long active core to reach the long equilibrium state for ideal CANDLE. Consequently, the modified CANDLE shape would be presented. The approximate characteristics of CANDLE burnup are observed in the finite and compact core, and the power density and fuel burnup are selected as main characteristic of modified CANDLE burnup. In this study, firstly, lots of optimization schemes were discussed, and one of optimization schemes was chosen at last to demonstrate the modified CANDLE burnup strategy. Secondly, for chosen compact small rector core, the modified CANDLE burnup strategy is applied and presented. Consequently, the new characteristics of this reactor core can be discovered both in ignition region and in fertile region. The results show that application of CANDLE burnup strategy to small modular PWR loading with FCM fuels suppresses the excess reactivity effectively and reduces the risk of small PWR reactivity-induced accidents during the whole core life, which makes the reactor control more safety and simple.

Investigation on thermo-mechanical performance of fully ceramic microencapsulated fuel

• A finite element method (FEM) model for FCM fuel thermo-mechanical performances analysis was presented. • The thermo-mechanical behaviors, coupled with irradiation behaviors were developed. • A 3D modeling of FCM fuel pellet with typical packing fractions was realized by an optimized stochastic strategy. • Effects of thermal boundaries, burnup, particle spacing, no fuel zone width and other key parameters were well investigated. Fully ceramic microencapsulated (FCM) fuel is a promising ATF concept with extraordinary fission product inclusion capacity, excellent oxidation and corrosion resistance, and high thermal conductivity. The thermo-mechanical performances are of great significance when evaluating the safety and applicability of a nuclear fuel. In this study, a finite element method (FEM) model for FCM thermo-mechanical performances analysis was presented. The thermo-mechanical behaviors, coupled with irradiation behaviors (including irradiation-induced dimensional change (IIDC), heat generation, conduction, thermal expansion, creep, burnup, and gap/plenum pressure, etc.), were developed to investigate the thermal and mechanical characteristics of TRISO-based FCM Fuel. A simplified model of TRISO fuel particles was established and validated. After that, a full 3D modeling of FCM fuel pellet with typical packing fractions was realized by an optimized stochastic strategy. A numerical method was validated and successfully applied to realize the simulations of a FCM fuel pellet. Two different thermal boundaries, namely fixed particle power and fixed pellet power, were performed to investigate the irradiation behavior of FCM fuel. Moreover, the effects of burnup, particle spacing, no fuel zone width and other key parameters were well studied and investigated. The performed calculation showed that temperature and stress in matrix and TRISO are both positively correlated with the packing fraction. It's worth noting that compared with the cases of fixed particle power, cases of fixed pellet power may lead to higher fuel temperature and higher stress. The results also implied that from the perspective of probability evaluation, most of the TRISO particles in FCM fuel can maintain the structural integrity under different packing fractions, but some of the particles with higher stress may crack or damage, especially in the cases of high burnup and high packing fraction.

Effective multi-group cross section calculation method for FCM fuel based on improved disadvantage factor method

Fully Ceramic Micro-encapsulated (FCM) fuel is an important candidate for the accident tolerant fuel (ATF). Compared with traditional fuel, the double heterogeneity of FCM fuel makes the effective multi-group cross section calculation more challenging. In this paper, an improved disadvantage factor method is proposed to deal with the self-shielding effect of FCM fuel in the resonance energy range and non-resonance energy range, to achieve the equivalent homogenization of the FCM fuel. A new equivalent particle–matrix model was constructed by using the particle Dancoff factor to overcome the problem that the traditional volume-weight equivalent model could not consider the macro heterogeneity between fuel rods. Based on the new one-dimensional equivalent sphere model, the ultrafine group slowing down equation is solved to obtain the ultrafine group disadvantage factor in the resonance energy region. In the non-resonance energy region, the multi-group disadvantage factors of the fast group and thermal group are obtained by using the eigenvalue calculation in the new equivalent particle–matrix model. The proposed method has been implemented in the high fidelity neutronics program NECP-X and tested with a set of cases. The results show its good agreement with the Monte Carlo reference for both the reactivity and self-shielding cross sections.

Development and application of Monte Carlo and COMSOL coupling code for neutronics/thermohydraulics coupled analysis

The tightly coupled solution of neutronics and thermal-hydraulics play an important role in precise simulation of fuel designs. Fully Ceramic Microencapsulated (FCM) fuel was one of the popular designs for accident tolerant fuel. To simulate advanced fuel types such as FCM fuel, a new method was proposed to couple the Monte Carlo code RMC with multi-physical software COMSOL. The MATLAB interface was adopted for coupling, and the superimposed structured mesh was applied to the mesh mapping and data interpolation between the structured meshes of RMC and unstructured meshes of COMSOL. Then the coupling code was verified by a typical PWR fuel rod and a FCM case. Results show that the calculated power and temperature distributions are reasonable, which can capture the self-shielding effect of thermal neutrons. Comparing with the traditional “cell-to-cell” mapping, the proposed method has advantages of good geometric flexibility and high accuracy, which can be applied to multi-physics simulations of different fuel designs.

Preliminary numerical investigation of TRISO-matrix interface debonding characteristics in fully ceramic microencapsulated fuel

The bonding mode of NITE-SiC matrix and TRISO particles has a great impact on the fracture mechanisms of the Fully Ceramic Microencapsulated (FCM) fuel. Combination strength and interface debonding characteristics are of great significance when evaluating the thermo-mechanical behavior of a FCM fuel. In this study, a finite element method (FEM) model for TRISO-matrix debonding characteristics analysis was presented based on the cohesive zone model (CZM). The CZM was coupled with several irradiation behaviors (burnup, fission gas release, gap/plenum pressure, etc.) and thermo-mechanical behaviors (irradiation-induced dimensional change (IIDC), contact, heat generation, conduction, gap heat transfer, thermal expansion, and creep etc.), and developed to investigate the interface debonding characteristics of TRISO coated FCM Fuel preliminarily. A 3-dimensional multiphysics modeling approach has been successfully applied to realize simulations of a representative FCM fuel unit, also a uniformly distributed FCM layer. The NITE-SiC and OPyC interface debonding characteristics model was well performed, and was validated with CZM theory. Moreover, the effects of burnup, particle spacing, stress distributions of matrix, interfacial bonding strength and other key parameters in CZM model, were studied and investigated.

Processing of fully ceramic microencapsulated fuels with a small amount of additives by hot-pressing

The effects of processing parameters on the properties of SiC matrices and fully ceramic microencapsulated (FCM) fuel pellets were investigated with a small amount of additives by hot pressing. The increases in the applied pressure as well as the sintering time were beneficial in improving the thermal conductivity of the SiC matrix and FCM pellets. Tristructural isotropic (TRISO) particles without an outer pyrolytic carbon layer are not suitable for processing FCM pellets because of the corrosion of the chemical-vapor-deposited SiC layer by the additive-derived liquid phase during sintering. The FCM pellet sintered with 2.71 wt% AlN–Y2O3–Sc2O3–MgO additives at 1870 °C for 4 h under 30 MPa in argon exhibited an excellent thermal conductivity (61.7 Wm−1 K−1 at a TRISO particle content of 43.3 %), owing to the (i) reduced amount of additives, (ii) high sintered density (residual porosity ∼ 0.4 %), and (iii) highly thermally conductive SiC matrix (105.0 Wm−1 K−1).

Transition cycle analysis of light water-cooled SMR core loaded with MOX (TRU) and FCM (TRU) fueled PWR assemblies

In this study, a light water-cooled small modular reactor (SMR) core loaded with special fuel assemblies composed of mixed oxide (MOX) fuel rods using UO2-TRUO2 and fully ceramic micro-encapsulated (FCM) fuel rods using TRUO2 was suggested and analyzed for transuranic (TRU) transmutation as an alternative option before the commercialization of sodium-cooled fast reactor (SFR). Especially, the FCM rods were loaded with only TRUO2 for deep burning of TRU. This study performed neutronic analysis of the transition cycle cores with the new fuel assemblies. In addition to the neutronic analysis, detailed analyses on two important issues for the light water-cooled reactor (LWR) core loaded with TRU were performed in fuel assembly level calculation. First, an effective burnable absorber design was searched to mitigate the degradation in effectiveness of the burnable absorber caused by the hardened neutron spectrum. Secondly, a reactivity decomposition analysis was performed for understanding quantitative nuclide-wise and reaction-wise contributions to the void reactivity which is a key safety parameter of TRU-bearing fuel assemblies. The analysis on the TRU mass flow through the transition cycles showed that a high net TRU consumption rate of 14.7% can be achieved at the equilibrium cycle.

METHODS FOR CALCULATION OF SELF-SHIELDED CROSS SECTIONS OF FULLY CERAMIC MICRO-ENCAPSULATED FUEL DOUBLE HETEROGENEOUS SYSTEM

Fully ceramic micro-encapsulated (FCM) fuel is a kind of fuel that employs tri-structural isotropic (TRISO) particles to enhance safety. The FCM fuel assembly is a double heterogeneous system. The conventional self-shielding calculation methods cannot treat the DH effect. In this paper, three methods based on equivalent homogenization of the TRISO particle and the matrix are studied and compared: the hyper-fine energy group cross sections (XSs) homogenization based hyper-fine energy group method (HHM), the hyper-fine energy group XSs homogenization based subgroup method (HSM) and the subgroup XSs homogenization based subgroup method (SSM). These methods are implemented in a high-fidelity neutronics code NECP-X. Numerical results show that these methods are able to treat the double heterogeneity of the FCM fuel. The precision of the HHM and HSM is higher than that of the SSM.

NEUTRONICS ANALYSIS OF VVER-1000 CORE USING AT-FCM FUEL

Using the Fully Ceramic Microencapsulated (FCM) fuel in light water reactors has multiple advantages, as it is accident tolerant because of; no hydrogen generation due to the cladding interaction with steam at high temperature, better retention of fission fragments and proliferation resistant due to very small production of transuranic elements during the burnup as compared to the standard UO2 fuel. In this study neutronics analysis of AT-FCM fuel consisting of TRISO particles embedded in SiC matrix is performed for replacement in existing VVER-1000 reactors. Standard VVER-1000 fuel assembly is transformed to Accident Tolerant Fully Ceramic Microencapsulated (AT-FCM) fuel assembly based on hydraulic diameter of the VVER-1000 assembly, the number of fuel pins are decreased with increased diameter and enrichment to conserve the initial fissile loading in AT-FCM assembly. Fuel centerline temperature of the AT-FCM assembly is found to be lower than the reference UO2 fuel assembly at the same total power produced because of the much higher thermal conductivity. FCM-TRISO fuel assembly namely Array 15 with 169 pins is proposed and analyzed. Pin cell, assembly level and full core calculations have been performed with SERPENT code using implicit and explicit models. VVER-1000 full core is modelled using the transformed FCM assembly. The embedded TRISO particles in a SiC matrix and the use of FeCrAl cladding turns out to be the perfect case for accident tolerance. High burnup of AT-FCM core in terms of MWd/kgHM for the same number of EFPDs is observed as compared to reference UO2 core due to the small breeding of transuranic elements Pu-239, Pu-240 and Pu-241. Appreciable quantity of the power is produced due to the fission of transuranic elements in reference UO2 assembly so the burnup in MWd/kgHM remains smaller than the AT-FCM fuel. Comparatively more softening of spectrum is found in AT-FCM fuel cells and assemblies towards the middle of the cycle (MOC) and End of the Cycle (EOC), this softening of spectrum tends to increase the rate of U-235 depletion. Very small quantities of plutonium isotopes are produced in AT-FCM as compared to the reference UO2 assembly because of small loading of U-238 at the BOC. The neutronics performance of AT-FCM core with burnable poison consisting of Gd2O3 and Er2O3 turn out to be better than reference UO2 assembly as it exhibits smooth burnup. Fuel Temperature Coefficient (FTC) and Moderator Temperature Coefficient (MTC) of the AT-FCM assembly is negative for most part of the cycle however, towards the end of cycle it becomes less negative due to small quantities of resonance absorbers, softening of thermal flux and increased rate of fission absorption in UO2.