Reactor Fuel Research Articles

Modernizing the active neutron collar poison rod correction using simulations to enhance PWR fresh fuel verification

The declared inventory of 235U in fresh, low enriched uranium, nuclear fuel assemblies is routinely verified by nuclear safeguards inspectorates using the UNCL (Uranium Neutron Collar – Light Water Reactor Fuel) nondestructive assay instrument. The trend in modern fuels is towards higher initial enrichment, which in turn requires a larger number of burnable poison pins with higher Gadolinia concentration to hold down the initial reactivity. UNCL assay error for modern fuel assemblies typically exceeds 10%. The traditional algorithm used to correct the response for burnable poison content needs revision to achieve accuracy comparable to the σR of 4.5% expected for poison-free assemblies. In this paper we review the problem and use a large set of 287 Monte Carlo simulations based on an experimentally benchmarked model corresponding to the 16×16 PWR array used in Brazilian Angra type II and III fuel. Simulated relative responses are used to evaluate the functional form of the poison correction and determine parameters optimal for 16×16 PWR assemblies. This update is conducted using between 4–24 Gadolinia pins ranging 2–11 wt% Gd2O3 with assemblies having mean enrichments ranging 2.5%–5%. Experimental validation of the updated coefficients using seven measurements shows bias is reduced by an order of magnitude and σR reduced from 10% to less than 2%. Two updated sets of coefficients are provided, the most accurate of which is directly useable in the existing analysis code (INCC) in use by inspectorate at fuel fabrication facilities.

Sensitivity of the active neutron coincidence collar response during simulated and experimental fresh fuel assay

Verification of the fissile (and fertile) content in fresh nuclear fuel assemblies is conducted by the IAEA to enforce the Nuclear Non-Proliferation Treaty using the UNCL (Uranium Neutron Collar — Light Water Reactor Fuel). The UNCL uses an uncorrelated AmLi neutron source to interrogate the fuel, producing a signal of coincident fission neutrons (Doubles rate) used to assay 235U content of the fuel. The cost of producing calibration assemblies and limited availability of diverse commercial assemblies at any one time historically restricted the ability to explore the full parameter space experimentally. Monte Carlo simulations can overcome this, but introduce additional sources of uncertainty. Here the sensitivity of simulations and measurements to various parameters is assessed for a reference 16×16 PWR assembly of uniform 3.2% enrichment. Uncertainty contributions in this evaluation include: simulated AmLi neutron source spectrum, AmLi neutron emission rate, AmLi anisotropicity, high density polyethylene (HDPE) density, the precise position of the fuel assembly within the detector, and experimentally the statistical uncertainty. The overall total systematic uncertainty estimate for the simulation of the absolute/relative Doubles rates responses are estimated to be approximately 2.0%/1.5%, and for experimental measurements systematic uncertainty reduces to 1.1%. This analysis supports further work using the relative Doubles rates in place of measurements for updating and extending the UNCL analysis methodology as systematic uncertainty is reasonably small.

Studying the Swirl Flow Hydrodynamics in the Guiding Channel Area Downstream of Mixing Spacer Grids of a PWR Reactor Fuel Assembly

The results from studying the coolant flow in fuel assemblies downstream of mixing spacer grids of a PWR reactor are presented. The aim of the study is to estimate the effect from using different designs of mixing grids. For this purpose, experimental investigations were carried out on an aerodynamic test facility with scaled models of fuel assembly fragments containing mixing spacer grids of different designs. Two adjacent cells of the guiding channel zone were selected as the representative study area. The specific structural feature of these cells is that the turbulizers installed on the intensifying grids have different spatial orientation. The general flow pattern is represented by the vector fields of tangential velocities and by the distribution of relative velocities in the gaps between the fuel rods and the guiding channel. For estimating the effect from using the grid designs with respect to coolant flow mixing, the intracell vortex formation coefficients and intercell mixing coefficients were analyzed. It has been shown from the performed analysis that, owing to the use of the alternative design with a pair of cell deflectors oriented in a changed direction, the intercell transfer coefficient and the intracell vortex formation coefficient were increased by a factor of 1.13 and 2.2, respectively, in comparison with those in the basic design. Hence, the use of the grid of the alternative design is more preferable for obtaining better mixing of the coolant flow. The accumulated database on coolant flow in the Square FA serves for engineering substantiation of the PWR reactor cores. The results from experimental studies are used for verification of CFD codes developed both in Russia and abroad and of programs for detailed cell-wise analysis of the cores for reducing the conservatism in substantiating their thermal reliability.

Effect of TiN Coating on the Fouling Behavior of Crud on Pressurized Water Reactor Fuel Cladding

A titanium nitride film was applied to mitigate the deposition of fouling deposits (crud) on pressurized water reactor fuel cladding. Titanium nitride was chosen as it reduces the van der Waals force, believed to be the dominant surface force at pressurized water reactor conditions, between crud and the substrate. Crud deposition experiments were conducted in a heated, flowing pressurized water reactor-simulant loop with uncoated and titanium nitride-coated Zr-Nb-Sn alloy tubes for 1–2 weeks. Crud morphology and thickness were investigated using focused ion beam–scanning electron microscopy, and crud chemical composition was measured via energy-dispersive X-ray spectroscopy. Results demonstrate that the titanium nitride coating effectively reduced crud deposition thickness compared to uncoated Zr-Nb-Sn alloy in the 2-week experiment. In addition, the titanium nitride coating increased the Ni/Fe ratio of the crud, which may help mitigate crud-induced power shifts.

Design and operation of a multi-stage reactor system for chemical looping combustion process

A chemical looping combustion (CLC) system with a multi-stage fuel reactor (FR) and a two-stage air reactor (AR), was constructed and operated in an experimental study and simulated via the computational particle fluid dynamics (CPFD) method. Hydrodynamic results with a bubble-based EMMS drag model agreed well with the experimental measurements. The wall effect in the small-scale AR led to an uneven particle distribution and irregular fluidization at a small NAR. The bubble size and velocity were limited in the FR, thereby significantly boosting the gas-solid contact. The particle residence time in the FR was prolonged, resulting in a high carbon capture rate. With the increase in fluidization number in FR and AR, the mass flux increased, but the residence time decreased. Adding a secondary gas flow in the FR has a positive impact on the solid circulation rate. The results in the thermal test indicated that a high CO2 capture rate and a low oxygen demand were achieved; thus the proposed reactor design improved the fuel conversion in the CLC process.

Application and Benchmarking of a Novel Coupled Methodology for Simulating the Thermomechanical Evolution of Sodium-cooled Fast Reactors Fuel Subassemblies

We employ a novel methodology to simulate the irradiation of Sodium cooled Fast Reactors (SFR) subassemblies. The proposed approach allows considering the coupling between key thermal–hydraulic and thermomechanical phenomena, which is usually neglected in traditional simulation methods. Comparisons between results obtained from coupled and uncoupled methodologies show that the latter can significantly over predict the stresses and the strains of highly irradiated fuel bundles. Additionally, we explore different options for the pin mechanical modeling. In particular, we show that a relatively simple beam-based finite element model for the fuel pins can be used to efficiently compute swelling and irradiation creep strains averaged over the cross section of the fuel claddings. However, if a prediction of the cladding swelling gradients is needed, the beam model must be replaced by a more computationally expensive 3D fuel pin model. An efficient method to estimate the fuel cladding temperature evolution based on a limited number of Computational Fluid Dynamics (CFD) simulations is also presented in this work, and its performance is evaluated. Finally, a numerical benchmark against the state-of-the-art coupled code system for the simulation of SFR subassemblies is presented. This benchmark shows that the proposed coupled methodology allows obtaining results that are in good agreement with the reference methodology. However, two clear advantages of the proposed methodology were identified. Firstly, the use of detailed CFD simulations in our methodology, compared to the lower resolution thermal-hydraulic model employed in the state-of-the-art approach. Secondly, considering the reduction of the coolant mass flow rate caused by the fuel bundle deformation, neglected in the pre-existing approach, which is shown in this work to have a potentially large impact on the coolant temperature distribution.

Radiation dose distribution of liquid fueled thorium molten salt reactor

A liquid fueled thorium molten salt reactor (TMSR-LF), one of the Generation IV reactors, was designed by the Shanghai Institute of Applied Physics, Chinese Academy of Sciences. This study uses the`rt code to calculate the neutron and gamma dose rate distributions around the reactor. Multiple types of tallies and variance reduction techniques were employed to reduce calculation time and obtain convergent calculation results. Based on the calculation and analysis results, the TMSR-LF1 radiation shield with a 60-cm serpentine concrete layer and a 120-cm ordinary concrete layer is able to meet radiation requirements. The gamma dose rate outside the reactor biological shield was 16.1 mSv h−1; this is higher than the neutron dose rate of 3.71 × 10–2 mSv h−1. The maximum thermal neutron flux density outside the reactor biological shield was 1.89 × 103 cm−2 s−1, which was below the 1 × 105 cm−2 s−1 limit.

Sensitivity Studies to Assess the Impact of Geometry and Operating/Boundary Condition Perturbations on Thermal-Hydraulic Behavior of Advanced Fuel Channels in Pressure Tube Heavy Water Reactors with Uranium and Thorium-Based Fuels

This paper reports the results of subchannel thermal-hydraulic studies (using the ASSERT-PV code) of the effects of variations and uncertainties in operating/boundary conditions and geometry on the predictions of pressure drop, dryout power, and dryout location for two types of advanced, nonconventional fuels in a pressure tube heavy water reactor (PT-HWR) fuel channel with 12 fuel bundles. The fuel bundles tested include a 37-element fuel bundle made with SEUO2 (1.2 wt% 235U/U), with a central fuel element made of ThO2, and 35-element fuel bundle made with (LEU,Th)O2, using 5 wt% 235U/U low-enriched uranium (LEU), 50 wt% LEUO2, and 50 wt% ThO2. Results indicate that for a range of flow conditions, the dryout power for the thorium-based 35-element fuel bundle is 10% to 26% higher than that for the uranium-based 37-element fuel bundle. Variation/uncertainty in the pressure tube diameter has the most significant impact on the pressure drop, dryout power, and dryout location. Results from these studies may have implications for the operations of PT-HWRs with advanced fuels, and further modifications may be desirable to further enhance thermal-hydraulic margins.

Spatiotemporal statistical characteristics of multiphase flow behaviors in fuel reactor for separated-gasification chemical looping combustion of solid fuel

Separated-gasification chemical looping combustion (SGCLC) has high conversion efficiency for solid fuels. Fuel reactor (FR) in SGCLC is designed as a gasifier (GR) and a connected reduction reactor (RR) with gas, sand and oxygen carrier (OC) phases. In order to clearly know the multiphase flow characteristics, a FR prototype is designed and the gas-solid hydrodynamics in FR are investigated using computational fluid dynamic (CFD) method. The gas velocity and OC flowrate are changed in simulations to study their effects on flow behaviors. Bubble probability in GR increases from 0 at bottom to 0.64 at 0.33 m while the gas volume fraction reaches 0.80 herein under fundamental condition. Spatiotemporal statistical results of the characteristics of gas and solid phases demonstrate that increasing gas velocity can enhance the heterogeneity of pressure and gas velocity distributions in GR. The mean values of pressure and gas velocity standard deviations increase from 13.71 Pa to 18.66 Pa and from 2.08 m/s to 2.55 m/s at 0.33 m when gas velocity accelerates from 1.8 m/s to 2.2 m/s. On the contrary, it has opposite effect on heterogeneity of pressure distribution in RR. Increasing OC flowrate from 0.6 kg/s to 0.8 kg/s can enlarge OC inventory in RR and enhance the standard deviation of pressure distribution in RR. The mean value of pressure standard deviations in RR increase from 3.12 Pa to 4.82 Pa at 1.85 m. The gas-OC mixing and contact can be improved for sufficient combustion reaction.

MATRIX EXPONENTIAL METHODS FOR PARALLEL COMPUTING OF ISOTOPIC DEPLETION AND SPECIES TRANSPORT FOR MOLTEN SALT REACTOR ANALYSIS

Matrix exponential methods have long been utilized for isotopic depletion in nuclear fuel calculations. In this paper we discuss the development of such methods in addition to species transport for liquid fueled molten salt reactors (MSRs). Conventional nuclear reactors work with fixed fuel assemblies in which fission products and fissile material do not transport throughout the core. Liquid fueled molten salt reactors work in a much different way, allowing for material to transport throughout the primary reactor loop. Because of this, fission product transport must be taken into account. The set of partial differential equations that apply are discretized into systems of first order ordinary differential equations (ODEs). The exact solution to the set of ODEs is herein being estimated using the matrix exponential method known as the Chebychev Rational Approximation Method (CRAM).

APPLICATION OF STUDSVIK’S CMS5 CODE SYSTEM TO ACCIDENT TOLERANT FUEL CORE DESIGN AND ANALYSIS

The possible deployment of Accident Tolerant Fuels (ATF) for currently-operating Light Water Reactors (LWR) has prompted interest in the use of Studsvik’s CMS5 code system to support the analysis of such advanced ATF core designs. Various ATF concepts have been proposed; for example, uranium silicide (U3Si2) fuel, together with iron-based (FeCrAl) cladding. The purpose of this work is to showcase the application of the CMS5 code system, which includes the CASMO5 advanced lattice physics code and the SIMULATE5 three-dimensional nodal simulator, to the analysis of a U3Si2/FeCrAl ATF concept. Given that the CMS5 code system was designed from inception to enable the analysis of advanced core designs, only minor changes to the CASMO5 lattice physics code and SIMULATE5 core simulator are necessary. The current CASMO5 586 energy-group nuclear data library provides all the necessary data to support the generation of homogenized data for downstream use by SIMULATE5 for ATF. The SIMULATE5 nodal code, which features a simplified fuel pin model, requires updating various thermophysical properties corresponding to the U3Si2/SiC ATF fuel and the gaseous conductance models. An equilibrium core for the Integral Inherently Safe (I2S) LWR design developed by the Georgia Institute of Technology was selected. The results of the CMS5 simulation were compared with those in the literature and were found to be in good agreement, giving us confidence that the CMS5 package can be used in the modeling of LWR systems with ATF technology.

EFFECT OF IODINE AND MOISTURE ON THE MICROSTRUCTURE OF ZIRCALOY-4 UNDER SERVICE CONDITION IN PHWR

Fuel failures are always a cause of concern in any nuclear reactors as it increases the manrem consumption of radiation workers. Although performance of the fuels in pressurized heavy water reactors is good, but still fuel failures occur occasionally. Post irradiation examination (PIE) of the failed fuel elements indicates internal hydriding, not deuteriding, as a major cause for the failures, although secondary deuteriding occurs and, in a few cases, failures are associated with defects in the end plug weld. The sources of hydrogen are either fuel pellets or the clad or the graphite coating. Restriction has been imposed on maximum content of total hydrogen in the fuel element to 1 mg to prevent hydriding of the Zircaloy clad tube. Accidental pick up of hydrogen occurs, which could lead to failure of the fuel bundles. Experimental investigations have been conducted to understand the individual effect of iodine and accidental pick up of moisture on the microstructure of Zircaloy-4 end cap welded samples with graphite coating. Results indicate that severe hydriding in Zircaloy-4 samples due to the existence of internal moisture in presence of graphite under service condition may result in fuel failure and justifies the findings of PIE.

Effects of Th doping on mechanical properties of U<sub>1–</sub><i><sub>x</sub></i>Th<i><sub>x</sub></i>O<sub>2</sub>: An atomistic simulation

Since thorium (Th) owns high conversion ratio in thermal neutron spectrum, high melting temperature, high thermal conductivity and good corrosion resistance in high-temperature water, it can be doped into UO<sub>2</sub> based fuel to initiate the fission reaction, and improve the physical properties of UO<sub>2</sub>. Owing to the challenging experimental conditions and technologies, molecular dynamics (MD) simulations are conducted to investigate the influences of Th doping on the mechanical properties of U<sub>1–</sub><i><sub>x</sub></i>Th<i><sub>x</sub></i>O<sub>2</sub>. The phase transition from initial fluorite structure to the metastable scrutinyite phase when loading along the [001] direction is observed, which accords well with the previous density functional theory calculations. However, if U<sub>1–</sub><i><sub>x</sub></i>Th<i><sub>x</sub></i>O<sub>2</sub> is loaded along the [111] direction, only brittle fracture is observed. It is found that both the elastic modulus and fracture stress decrease linearly with elevating temperature but the fracture strain increases. As the Th concentration increases from 0 to 0.55, the elastic modulus first decreases and then increases; if the Th concentration is larger than 0.1, the fracture stress increases and the fracture strain decreases monotonically. The cracks are nucleated with an angle of 45º to the loading direction, propagate rapidly, and are characteristic of brittle fracture, which accords well with the classical failure criteria and experimental results for brittle materials. By comparison, the uniaxial tensile loading is also performed for polycrystalline U<sub>1–</sub><i><sub>x</sub></i>Th<i><sub>x</sub></i>O<sub>2</sub>. It is found that the elastic modulus and fracture stress decrease as the temperature increases. However, the elastic modulus is not sensitive to the Th concentration and the fracture increases as the Th concentration increases. The brittle intergranular fracture is observed in each of all polycrystalline samples. The obtained physical parameters are useful for designing the fuels in nuclear reactors.

VALIDATION OF COOLANT THERMAL RESPONSE IN A TRANSIENT FINITE DIFFERENCE THERMAL TRANSPORT MODEL WITH APPLICATIONS TO FAST SPECTRUM REACTORS

Transient behavior in nuclear reactors is important in accidents and with reactivity control systems that are driven by thermal feedback. Here, we describe a transient finite difference model for a pin cell system. The fidelity of the model is shown by validation against the thermocouple measurements of the CABRI BI1 experiment and the Safety Analysis System-Sodium Fast Reactor model of the experiment. In the BI1 experiment, a sodium-cooled mixed oxide fuel pin was subject to a loss of flow transient to coolant boiling within a sodium test loop positioned in the center of the CABRI research reactor. Comparisons to the initial steady-state coolant temperature profile, coolant temperature profile at twenty seconds into the transient, and at four axial locations within the coolant show agreement of the simple model with the experimental results better than or similar to those of the Safety Analysis System-Sodium Fast Reactor model. The model can be used to determine the thermal response times of coolant in fast reactors currently operating or in the design phase when subject to loss of flow accidents or other transients. Here, we investigate the difference in coolant thermal response for metal fueled and mixed oxide fueled sodium fast reactors when subject to transient overpower and loss of flow events. Additionally, we determine the effect of pin pitch on outlet coolant temperatures during the overpower event. Finally, we return to the CABRI experiment and show the importance of porosity in fuel temperature calculations.

CORE DESIGN ACTIVITIES OF THE VERSATILE TEST REACTOR – CONCEPTUAL PHASE

The Versatile Test Reactor (VTR) is a new fast spectrum test reactor being developed in the United States under the direction of the US Department of Energy, Office of Nuclear Energy. The VTR mission is to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. This includes neutron irradiation capabilities which would support alternate coolants including molten salt, lead/lead-bismuth eutectic mixture, gas, and sodium. The VTR aims at addressing most of the needs of the various stakeholders, which is primarily composed of advanced reactor technologists, developers and vendors, as well as a number of others interested parties.Design activities are underway targeting a first criticality date by 2026, with General Electric recently joining the project to contribute to the VTR plant design. Current efforts are focused on all aspects of the VTR design, with the core design being at the center of the initial steps. The VTR is currently proposed as a 300 MWth sodium-cooled fast reactor able to reach peak fast flux levels in excess of 4.0x1015 n/cm2-s (and total flux level of about 6.0x1015 n/cm2-s). In this configuration, it is using ternary metallic fuel with reactor-grade plutonium and 5% low-enriched uranium.

UNCERTAINTY QUANTIFICATION IN STEADY STATE SIMULATIONS OF A MOLTEN SALT SYSTEM USING POLYNOMIAL CHAOS EXPANSION ANALYSIS

Uncertainty Quantification (UQ) of numerical simulations is highly relevant in the study and design of complex systems. Among the various approaches available, Polynomial Chaos Expansion (PCE) analysis has recently attracted great interest. It belongs to nonintrusive spectral projection methods and consists of constructing system responses as polynomial functions of the stochastic inputs. The limited number of required model evaluations and the possibility to apply it to codes without any modification make this technique extremely attractive. In this work, we propose the use of PCE to perform UQ of complex, multi-physics models for liquid fueled reactors, addressing key design aspects of neutronics and thermal fluid dynamics. Our PCE approach uses Smolyak sparse grids designed to estimate the PCE coefficients. To test its potential, the PCE method was applied to a 2D problem representative of the Molten Salt Fast Reactor physics. An in-house multi-physics tool constitutes the reference model. The studied responses are the maximum temperature and the effective multiplication factor. Results, validated by comparison with the reference model on 103 Monte-Carlo sampled points, prove the effectiveness of our PCE approach in assessing uncertainties of complex coupled models.

Numerical simulation of commercial scale autothermal chemical looping reforming and bi-reforming for syngas production

Autothermal Chemical Looping Reforming (a-CLR) is an emerging technology that facilitates CO2 capture and minimizes energy losses in syngas production. The dual-fluidized bed process uses a bubbling fuel reactor (FR) and a riser air reactor (AR). A 1-D multiscale model was developed that couples fluidized bed hydrodynamics with intrinsic reaction kinetics, including catalyst deactivation by oxidation and coke formation. An a-CLR unit with a capacity equivalent to 50 conventional reformer tubes was considered to demonstrate feasibility. The effects of the catalyst activity, main operation conditions and CO2 co-feeding on the performance were then analyzed. The results confirm a-CLR is feasible with realistic reactors dimensions and solids circulation rate. To avoid the risk of sintering and catalyst deactivation, the oxygen carrier should only be slightly oxidized in the AR. Autothermal operation then requires an oxygen-to-CH4 feed ratio of around 1.2. The high catalyst-to-gas feed ratio in the FR and bubble-emulsion phase mass transfer limitations result in a low sensitivity of the methane conversion to the catalyst activity. A low H2O-to-CH4 feed ratio introduces a risk for coke formation in the bottom of the FR, but part of the feed methane is fully oxidized, producing H2O and CO2 in the emulsion phase and mitigating this risk. Co-feeding CO2 with CH4 and H2O is seen to allow adjusting the H2/CO-ratio of the syngas, but increases the risk of coke formation. Finally, heat recovery from the flue gas and syngas from both reactors by preheating the feed gases further increases the energy efficiency.

A polyester-based initiation strategy for achieving high-efficient cracking of hydrocarbon fuels

Active cooling of scramjets by endothermic hydrocarbon fuels is considered as an effective approach to addressing the thermal problem for hypersonic aircrafts. To accelerate the endothermic cracking reactions of these fuels for a better cooling capability, a strategy for hydrocarbon cracking using polymer-based macromolecular initiators has been proposed. To verify the universality of this strategy and also to enhance the solubility of macromolecular initiators, a novel polyester macromolecular initiator (MI) is designed. It is clarified that the MI features high-efficient thermal cracking of hydrocarbon fuels in a tubular reactor with an improved conversion ratio, gas yield, and alkene selectivity. The possible mechanism behind the macromolecular initiator is revealed on the basis of intermediate analyses and density functional theory (DFT) calculations, which show the radical fragments from MI decomposition are effective in accelerating JP-10 dehydrogenation. The successful implementation of polyester materials in hydrocarbon fuels holds immense promise for the optimal use of hydrocarbon fuels in hypersonic aircrafts.

Exergy analysis and dynamic control of chemical looping combustion for power generation system

Chemical looping combustion (CLC) is an efficient technology for power generation system (PGS) and CO2 capture. Regarding carbon deposition, this work presents a global energy optimization strategy by integrating CLC with PGS process (CLC-PGS) based on exergy analysis. First, the CLC-PGS system is simulated by using Aspen Plus software and its steady-state simulation results are used for global exergy analysis. Second, the carbon deposition phenomena resulting in insufficient contact between oxygen carrier (OC) and methane (CH4) and thus low conversion of CH4 is considered in the CLC-PGS optimization. OC circulation rate is increased to address carbon deposition problem and alleviate exergy destruction along with fuel reactor (FR) temperature, the ratio of CH4 to steam (XCTS), and the ratio of CH4 to OC (XCTO) optimized simultaneously by exergy analysis. Finally, the pressure and liquid level of drum in heat recover steam generator (HRSG) are selected as key variables by transfer entropy method for control scheme design. The dynamic behaviors of three alternative control schemes are compared by Aspen Dynamics software, with the control scheme II resulting the best according to the stability indicators in the face of ±10% disturbances.

A new burn-up module for application in fuel performance calculations targeting the helium production rate in (U,Pu)O2 for fast reactors

In light of the importance of helium production in influencing the behaviour of fast reactor fuels, in this work we present a burn-up module with the objective to calculate the production of helium in both in-pile and out-of-pile conditions tracking the evolution of 23 alpha-decaying actinides. This burn-up module relies on average microscopic cross-section look-up tables generated via SERPENT high-fidelity calculations and involves the solution of the system of Bateman equations for the selected set of actinide nuclides. The results of the burn-up module are verified in terms of evolution of actinide and helium concentrations by comparing them with the high-fidelity ones from SERPENT, considering two representative test cases of (U,Pu)O2 fuel in fast reactor conditions. In addition, a code-to-code comparison is made with the independent state-of-the-art module TUBRNP (implemented in the TRANSURANUS fuel performance code) for the same test cases. The herein presented burn-up module is available in the SCIANTIX code, designed for coupling with fuel performance codes.