Generation Nuclear Systems Research Articles

Design, fabrication and validation of an electrical conductivity principle based two-phase detection sensor array for molten lead (Pb) based heavy metal coolants up to 600°C

Molten lead (Pb) and its alloys (PbBi and PbLi) are of immense interest for various nuclear engineering applications, including but not limited to advanced Lead-cooled Fast Reactors (LFRs), tritium Breeding Blankets (BBs) of fusion power plants and spallation targets for Accelerator-Driven Systems (ADS). Owing to their attractive thermophysical properties, these advanced fluids assert their candidacy to address the critical requirements of neutron multiplication, neutron moderation, high temperature coolants and tritium breeders, enabling the operation of next generation nuclear systems at high temperatures with better efficiencies. However, for numerous reasons such as a compromise of structural integrity at the heat transfer interface, presence of an inert cover gas during charging of molten metal in the loop, and the fusion fuel cycle itself may lead to molten metal-gas two-phase flows with high density ratios. At present, no effective diagnostics exist to detect such operational and accidental occurrences in high temperature molten metal systems resulting in a severe lack of relevant experimental studies. To address these limitations and to advance the current understanding toward two-phase regimes in high temperature Pb-based melts, the present work focuses on the design and assembly aspects of an electrical conductivity-based two-phase detection sensor array, utilizing high purity α-Al2O3 coatings with AlPO4 binder as electrical insulation layers. This paper discusses the design considerations, thermal analysis, systematic selection of structural/functional components along with preliminary results from the probe performance tests in very high temperature (600°C) static molten Pb column for real time detection of argon gas bubbles rising within the melt. Quantitative estimations of time-averaged void fraction, average bubble impaction frequency and average bubble residence time are presented from the preliminary experimental investigations.

Development of Fe–Cr–Si deposited layer manufactured by laser directed energy deposition process

In this study, the mechanical and corrosion characteristics of a corrosion-resistant layer made of stainless steel (STS) 316 L and Fe–Cr–Si alloy powder were investigated using laser-directed energy deposition (DED). In the STS 316 L deposited specimen, both the substrate and deposited layer were face-centred cubic (FCC). The deposited Fe–Cr–Si layer was clearly separated from the substrate because it was composed of body-centred cubic (BCC). Despite the phase differences, the surface of the Fe–Cr–Si-deposited layer showed a lower corrosion rate than that of the STS 316 L. All the deposited specimens exhibited typical high-temperature tensile behavior. However, the Fe–Cr–Si deposited layer at 600 °C showed a notable reduction in strength and increased elongation compared to the room temperature (RT) and 300 °C test results owing to the carbide concentration and phase transformation in the deposited layer. Because nuclear facilities mainly operate at temperatures below 600 °C, Fe–Cr–Si materials can also be used as nuclear piping coating materials. This study provides a mechanism for the high-temperature properties and corrosion resistance of the Fe–Cr–Si deposited layer and makes it competitive for application in fourth generation nuclear power systems.

Research and analysis of the thermal and control characteristics of the plate-type fuel assembly for the supercritical CO2 reactor

Small nuclear reactors can be applied to micro grid power generation, island and space nuclear power systems. The plate-type fuel assembly has a compact design structure, low fuel core temperature, large heat exchange area, and high heat exchange efficiency, which is widely used in small nuclear reactors. To supplement the lack of research in S-CO2 small nuclear reactor, The Brayton cycle reactor system analysis program of S-CO2 plate-type fuel assemblies (BRESA-PFA) is developed, and the reactor control system program is developed to simulate the transient condition in this paper. This article establishes a fuel assembly flow and heat transfer model and a control rod control model using Modelica language, achieving physical and thermal coupling and power control of the reactor. The steady-state operating parameters of the developed program BRESA-PFA were studied, and the flow distribution, coolant and fuel temperature distribution of BRESA-PFA were obtained. The flow distribution conforms to the experimental parameters of CARR, and the fuel assembly radial temperature distribution is high at the center and low at the edge. After verification, the maximum fuel temperature did not exceed the limit value and met the design requirements. Transient characteristic analysis was conducted to study the rod lifting strategy during the startup process, the start-up and rod lifting process of BRESA-PFA is N2-N1-G2-G1, using the logic of step control. Research on reactor control system response under loss of coolant accident, after the accident, the coolant flow suddenly decreased to 65 % of the rated value, and the reactor power also decreased to 65 % FP, G2 and G1 control rods were inserted downwards, the coolant temperature fluctuates by 1 %∼2%. The reactor control system was able to respond quickly, ensuring the stability of coolant temperature, proving the safety and reliability of BRESA-PFA, and providing theoretical reference and scheme support for the research of S-CO2 small nuclear reactors.

Comparative study of neutronic, mechanical and thermodynamic properties of accident tolerant cladding materials: SiC, TiC and ZrC

In the framework of the using carbide materials as alternative nuclear fuel cladding material, an investigation has been performed to obtain neutronic, mechanical and thermodynamic behaviors under different conditions and high temperature. To achieve this goal, we have used neutron transport theory methods to study neutronic performance and density functional theory methods for mechanical and thermodynamic kinetics of carbides, SiC, ZrC and TiC. In the neutronic calculations, we have compared the results of carbides with widely used cladding materials, such as baseline Zircaloy, 304SS, 310SS, FeCrAl and APMT. On the other hand, the carbides have been analyzed from the perspective of mechanical and thermodynamic properties. The obtained lattice parameters, elastic constants, elastic modulus, Poisson’s ratio, hardness, heat capacity, Debye temperature distribution, enthalpy, free energy and entropy have been discussed. The presented results advance underlying mechanisms of SiC, TiC and ZrC and guide future research, such as fuel-cladding systems, structural materials and new generation nuclear reactor systems.

Evaluation of radiation resistance of an austenitic stainless steel with nanosized carbide precipitates using heavy ion irradiation at 200 dpa

Despite many advantages as structural materials, austenitic stainless steels (SSs) have been avoided in many next generation nuclear systems due to poor void swelling resistance. In this paper, we report the results of heavy ion irradiation to the recently developed advanced radiation resistant austenitic SS (ARES-6P) with nanosized NbC precipitates. Heavy ion irradiation was performed at high temperatures (500 °C and 575 °C) to the damage level of ∼200 displacement per atom (dpa). The measured void swelling of ARES-6P was 2–3%, which was considerably less compared to commercial 316 SS and comparable to ferritic martensitic steels. In addition, increment of hardness measured by nano-indentation was much smaller for ARES-6P compared to 316 SS. Though some nanosized NbC precipitates were dissociated under relatively high dose rate (∼5.0 × 10−4 dpa/s), sufficient number of NbC precipitates remained to act as sink sites for the point defects, resulting in such superior radiation resistance.

Materials for Sustainable Nuclear Energy: A European Strategic Research and Innovation Agenda for All Reactor Generations

Nuclear energy is presently the single major low-carbon electricity source in Europe and is overall expected to maintain (perhaps eventually even increase) its current installed power from now to 2045. Long-term operation (LTO) is a reality in essentially all nuclear European countries, even when planning to phase out. New builds are planned. Moreover, several European countries, including non-nuclear or phasing out ones, have interests in next generation nuclear systems. In this framework, materials and material science play a crucial role towards safer, more efficient, more economical and overall more sustainable nuclear energy. This paper proposes a research agenda that combines modern digital technologies with materials science practices to pursue a change of paradigm that promotes innovation, equally serving the different nuclear energy interests and positions throughout Europe. This paper chooses to overview structural and fuel materials used in current generation reactors, as well as their wider spectrum for next generation reactors, summarising the relevant issues. Next, it describes the materials science approaches that are common to any nuclear materials (including classes that are not addressed here, such as concrete, polymers and functional materials), identifying for each of them a research agenda goal. It is concluded that among these goals are the development of structured materials qualification test-beds and materials acceleration platforms (MAPs) for materials that operate under harsh conditions. Another goal is the development of multi-parameter-based approaches for materials health monitoring based on different non-destructive examination and testing (NDE&T) techniques. Hybrid models that suitably combine physics-based and data-driven approaches for materials behaviour prediction can valuably support these developments, together with the creation and population of a centralised, “smart” database for nuclear materials.

Numerical research on thermal stress of steam cooler tube sheet mechanical structure

Steam cooler is one of the most important mechanical equipment in thermal power generation system and nuclear power generation system. The steam cooler bears a huge temperature gradient load when the working conditions are switched. In order to analyze the thermal stress of steam cooler tube sheet with high temperature load under harsh working conditions, the thermal-structural coupling analysis model of steam cooler tube sheet is constructed with finite element method. The results show that the stress concentration exists at the position connection between heat exchanger tube and cylinder. The maximum stress is located at the outermost heat exchanger tube with the peak stress of 320MPa. The heat exchanger tube layout alone cause higher stress. This situation should be avoided in the design of heat exchanger tube sheet. Furthermore, the strength and safety of the steam cooler tube sheet are evaluated with the stress linearization method. The steam cooler tube sheet design meets the safety requirements of structural strength under high temperature load.

Effects of irradiation defects on the adsorption of oxygen on 3C-SiC low index surfaces

Silicon carbide (SiC) is widely regarded as a potential accident tolerant fuel cladding material for the next generation nuclear power system. Oxidation is the leading cause of failure of SiC in a high temperature environment. Adsorption of oxygen on SiC surface is the first stage of SiC oxidation. Effects of irradiation defects, such as vacancy and antisite atom, on the adsorption of oxygen on 3C-SiC low index surfaces are investigated by first principles calculation method. The results show that the oxygen molecule tends to dissociate spontaneously into two O atoms when adsorbed on the low index surfaces. The adsorption stability of oxygen molecule on the Si-terminated surface is higher than that on the C-terminated surface. As a whole, irradiation defects decrease the stability of oxygen adsorption on C-(100) and Si-(100) surfaces but increase the stability of oxygen adsorption on (110), (111), and (1¯1¯1¯) surfaces. Except for (110) surface, new covalent bonds as the precursors of oxides are formed on the SiC surfaces due to the irradiation defects, which changed the oxide species of other surfaces during the initial oxidation. Bader charge analysis and the local density of state results show that charge transfer plays a crucial role in the adsorption process.

Applications of carbide ceramics in nuclear reactors

<p indent=0mm>After the Fukushima nuclear accident in Japan, the safety requirements of nuclear energy systems are constantly increasing. A series of new nuclear energy systems are under development, such as very high-temperature gas-cooled reactor (VHTR), lead-cooled fast reactor (LFR), gas-cooled fast reactor (GFR) and so on. In order to achieve the goals of higher safety, these nuclear systems have put forward higher requirements for nuclear materials. For example, materials working in VHTR and GFR are supposed to withstand higher temperatures, and LFR requires materials to be more corrosion-resistant. Nuclear materials have to meet these requirements under normal operating conditions, as well as in incidental and accidental conditions. Therefore, it is necessary to optimize the existing material system and develop new high-performance materials to meet the future development needs of advanced nuclear power systems. Compared with traditional nuclear materials, carbide ceramics have better mechanical properties under elevated temperatures, stronger irradiation resistance and better comprehensive thermophysical properties. Up to now, great breakthroughs have been made in the research of carbide ceramic materials and some key materials are going to be used in the new-generation nuclear systems. For example, uranium carbide (UC) is gradually replacing traditional uranium dioxide (UO₂), and silicon carbide (SiC) is acting as the main fission product barrier layer of coated fuel particles in HTGR. Other new carbide materials such as zirconium carbide (ZrC) and MAX phase carbide materials are also becoming a new research hotspot in the field of nuclear materials. Compared with other ceramic materials, the fabrication especially its densification of carbide is generally more difficult. In addition, high purity and good neutron properties are also the basic requirements of carbide ceramic materials for nuclear energy. The vapor deposition process and powder metallurgy process have been used to fabricate carbide ceramics for different components in nuclear reactors. Some new preparation methods have been developed to fabricate fuels and structural materials. For example, gelation progress is applied to prepare UCO microsphere fuel, and nano infiltration and transient eutectic (NITE) process is used in the preparation of SiC_f/SiC composites, which is potential for nuclear claddings. The irradiation-induced degradation of nuclear ceramic materials in the reactor is unavoidable. In order to evaluate the performance changes of materials under irradiation conditions, different irradiation methods such as electron irradiation, ion irradiation, proton irradiation and neutron irradiation are used. Due to its unique microstructure and properties, carbide ceramics exhibit different properties from other nuclear materials under irradiation. The performance changes of main nuclear carbide materials after irradiation are summarized, including microstructure evolution, macroscopic irradiation-induced swelling, changes in mechanical and thermal properties, etc. In general, based on the current research progress, the article reviews the application of carbide ceramics in new nuclear power systems and elaborates their service scenarios. Basic properties, preparation methods and irradiation properties of carbide ceramic materials are introduced in detail. The future development directions and the processing techniques are proposed. In the future, carbide ceramic materials should be further developed, and carbide ceramic materials will find diverse applications in the new generation nuclear systems.

Performance of 14Cr ODS-FeCrAl Cladding Tube for Accident Tolerant Fuel

In the framework of Accident tolerant fuel (ATF) program, several types of claddings and pellets with enhanced accident tolerance have been developed for light water reactors. Oxide dispersion strengthened (ODS) FeCrAl alloys have been considered as a promising candidate for cladding materials due to their good mechanical strength, excellent structural stability and chemical durability at high temperature. The out-of-pile performance of 14Cr ODS-FeCrAl cladding tube fabricated by cold-rolling, such as microstructure, thermophysical property, mechanical property, and corrosion resistance, has been examined and discussed. The results confirm that iron-based ODS alloy is one of the promising candidates to be used as ATF cladding. It could also aid in the supplement of property database of ODS-FeCrAl for future use in nuclear cladding and structural applications in next generation nuclear systems.

A brief review of the development of high temperature gas cooled reactor

With the continuous development of nuclear energy, the high temperature gas cooled reactor (HTGR) is considered as one of the advanced reactor types of the fourth generation nuclear power system because of its new inherent safety concept, modular concept, high economic competitiveness and increasing construction flexibility. In this paper, the development history of high temperature HTGR is briefly summarized and the latest research is summarized.

Understanding bubble and void nucleation in dual ion irradiated T91 steel using single parameter experiments

Ferritic-martensitic steels are attractive candidates for structural materials in next generation nuclear reactor systems due to their resistance to radiation induced swelling. Cavity and dislocation loop evolution was characterized in dual ion irradiated T91 steel in three separate irradiation campaigns examining single parameter dependencies of temperature, helium co-injection rate, and damage rate. Irradiations resulted in bimodal cavity size distributions across nearly all ranges of experimental parameters. It was determined that irradiation temperature and helium co-injection rate are stronger influences on bubble stability and the transition from bubbles to voids than is the irradiation damage rate. At low helium injection rates all helium is in vacancy clusters that evolve into bubbles or voids. At high helium injection rates, bubbles become saturated with helium resulting in accumulation of helium at other traps such as dislocation loops. At intermediate levels of He that should aid in the nucleation of bubbles and enhance swelling, the high density of sinks in the F-M microstructure suppresses bubble nucleation and therefore, the onset of swelling. At high enough temperatures, helium is only in bubbles as other strong helium traps, such as dislocation loops, did not form. The mechanism of bubble to void transition was found to shift from being driven by the accumulation of helium to the critical bubble at low damage rates to being driven by spontaneous formation by stochastic vacancy fluctuation at high damage rates.

Light and heat-shock mediated TDA1 overexpression as a tool for controlled high-yield recombinant protein production in Chlamydomonas reinhardtii chloroplasts

The microalga Chlamydomonas reinhardtii offers a rapid, scalable and low-cost platform for recombinant protein production. Its chloroplast provides a particularly robust expression system for high yield production of complex proteins requiring challenging post-transcriptional modifications. Controlled transgene expression provides an important advantage supporting many applications. Currently, however, only three inducible systems exist for algal chloroplast expression regulation. Here, we present the nuclear encoded translation enhancer TDA1 that regulates transgene expression in the chloroplast, as a component of an inducible system that is designed for high-yield production. Specifically, the C-terminus of TDA1 (cTDA1) promotes translation of the atpA transcript by interacting with its 5′-UTR, which has been shown to support high levels of recombinant protein expression. cTDA1 under the control of the inducible promoter HSP70A-RBCS2 was introduced into a chloroplast mutant expressing GFP under the control of the atpA 5′-UTR. These cTDA1/GFP mutants were grown under optimised illumination regimes and subjected to specific heat-shock treatments, after which cTDA1 and GFP expression levels were measured. A ~1.9-fold increase was detected in induced samples grown under optimised production conditions (4 days cultivation period under 6 h·d−1 illumination at 200 μE·m−2·s−1 followed by 2 h light exposure (200 μE·m−2·s−1) after the end of the 4 day period; 30 min heat-shock at 40 °C with subsequent 5 h incubation at RT (200 μE·m−2·s−1 illumination)). Western Blot analysis confirmed that after induction, the cTDA1 and GFP accumulation correlated. This demonstrates proof-of-concept that expression of the nuclear cTDA1 regulates atpA 5′-UTR mediated chloroplast transgene expression and thus provides the basis for a next generation nuclear inducible chloroplast regulation systems.

Preliminary results on the study of 237Np(n,f) at n_TOF/EAR2 in energies related to nuclear waste transmutation

The accurate knowledge of neutron-induced fission cross sections of isotopes involved in the nuclear fuel cycle is essential for the optimum design and safe operation of next generation nuclear systems. Such experimental data can additionally provide constraints for the adjustment of nuclear model parameters used in the evaluation process, resulting in a further understanding of the nuclear fission process. In this respect measurements of the 237Np(n,f) cross section have been performed at the n_TOF facility at CERN in the horizontal 185 m flight-path (EAR1) which were discrepant by 7% in the MeV region. The neutron-induced fission cross section of 237Np(n,f) was recently restudied at the EAR2 19.5 m vertical beam-line at CERN’s n_TOF facility, over a wide range of neutron energies, from 100 keV up to 15 MeV, using the time-of-flight technique and a modern set-up based on Micromegas detectors. This study was performed in an attempt to resolve the aforementioned discrepancies and to provide accurate data of a reaction that is frequently used as reference in measurements related to feasibility and design studies of advanced nuclear systems. Preliminary results with a high statistical accuracy that resolve the discrepancies will be presented along with a brief discussion concerning the facility and the analysis.

The Basic Principles of Assessment of the Structural Integrity and Lifetime of the BN-Type Fast Neutron Reactor Components Considering Material Degradation

The paper presents an overview of the basic principles of extending the lifetime of BN-600 fast reactors. These principles based on analysis of the main in-service mechanisms of material embrittlement and damage underlie the normative documents of the Rosatom State Corporation and were used by the authors in their developments for justification of the design lifetime of the BN-800 and BN-1200 fast reactors. The so-called critical events and limit states that determine the structural integrity and lifetime of fast reactor components have been formulated. On the basis of the results of this work, the repost “Basic Principles for Lifetime and Structural Integrity Assessment of BN-600 and BN-800 Fast Reactor Components with Regard to Material Degradation” was made at the International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17) held by the IAEA in Yekaterinburg in 2017.

Optical fibre void fraction detection for liquid metal fast neutron reactors

Detection and characterisation of voids in heavy liquid metals (HLM) is required for next generation nuclear reactor systems. However, its determination presents a challenge owing to their opaque nature and the high temperatures involved. Therefore, tools are needed that can be used to capture local, quantitative information at relevant nuclear operating conditions.In this work, the feasibility of using optical fibre sensors for the measurement of void fractions in liquid metals is presented. Since the functioning of optical probes is usually explained by a crude on-off model, the complexity of both the optical response and the hydrodynamic tip–interface interactions often remain hidden to new users. To clarify this point, well-controlled lab-scale experiments dealing with the response of three concept probe tip geometries are presented. Analysis of the obtained signal transients is used to provide guidelines for effective data processing. To conclusively demonstrate the principle, an optimal prototype system successfully determined the local void fraction and frequency in a pilot-scale reactor using liquid lead bismuth eutectic (LBE) as working fluid.The information obtained by the optical sensor allows for validation of computational fluid dynamics (CFD) models to aid the design of LBE-based nuclear reactors, as well as for other liquid metal–gas systems in general.

Multifunctional nanoceramic coatings for future generation nuclear systems

Several breeding blanket concepts for the DEMO reactor employ the eutectic Pb–16Li as breeder material, namely Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL). These three concepts share, with different incidences, three major technological challenges: Tritium containment, steel corrosion and magnetohydrodynamic drag. Here, we describe the ongoing work on multifunctional Al2O3 nanoceramic coatings grown by Pulsed Laser Deposition (PLD) on Reduction-Activation Ferritic-Martensitic (RAFM) steel. In fact, PLD can produce relatively thick (up to tens of μm) high performance coatings with metal-like mechanical properties. The coatings were tested as Tritium permeation barriers with Hydrogen at different temperature (from 350 to 650 °C). Results collected in this way indicate an excellent behavior, with a permeation reduction factor (PRF) close to 105. In addition, it was shown that they are able to maintain similar properties even when Deuterium was employed and also under 2 MeVelectrons irradiation. [1] Moreover, the electrical conductivity of these dielectric coatings was shown to be extremely low even under irradiation. Finally, to evaluate the chemical compatibility of Al2O3 films in liquid eutectic Pb-16Li, PLD deposited samples have been exposed to static corrosion tests up to 8000 h. No corrosive attacks of the steel substrate are detected. To conclude, Alumina coatings deposited by PLD show optimal characteristics in order to tackle the major technological challenges associated to the Breeding Blanket (BB) concepts employing Pb-16Li as breeder materials.

Research on the structure design of the LBE reactor coolant pump in the lead base heap

Abstract Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.

Development of a sub-channel thermal hydraulic analysis code and its application to lead cooled fast reactor

The most important candidate for 4th generation nuclear system is lead or lead alloy cooled fast reactor. In the design phase of a lead cooled reactor, it is a top priority to finish the analysis work for the core. The detailed sub-channel analysis code KMC-Sub (Keda multi-physics and multi-scale coupling platform) has been developed to analyze steady state thermal hydraulic issues of SNCLFR-100 (Small Modular Lead-cooled Natural Circulation Fast Reactor, which was designed by University of Science and Technology of China). The model used in this code had taken the influence of cross flow into account, both forced and natural circulation can be simulated, to assess the development status of KMC-sub, experimental data from ORNL 19 pin tests (sodium cooled) and CAS 61 rods test (lead bismuth eutectic cooled) are compared to results from the code. The author found that in most flow rate and power density regime, the results coincides well with the tests data. After the V&V work, the code was used to analyze the flow and temperature distribution in important assemblies of SNCLFR-100 core, which showed that the design is reasonable and feasible.

Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like sodium fast reactor

The EU 7th Framework ESNII+ project was launched in 2013 with the strategic orientation of preparing ESNII for Horizon 2020. ESNII stands for the European Industrial Initiative on Nuclear Energy, created by the European Commission in 2010 to promote the development of a new generation of nuclear systems in order to provide a sustainable solution to cope with Europe’s growing energy needs while meeting the greenhouse gas emissions reduction target. The designs selected by the ESNII+ project are technological demonstrators of Generation-IV systems. The prototype for the sodium cooled fast reactor technology is ASTRID (standing for Advanced Sodium Technological Reactor for Industrial Demonstration), which detailed design phase is foreseen to be initiated in 2019. The ASTRID core has a peculiar design which was created in order to tackle the main neutronic challenge of sodium cooled fast reactors: the inherent overall positive reactivity feedback in case of sodium voiding occurring in the core. Indeed, the core is claimed by its designers to have an overall negative reactivity feedback in this scenario. This feature was demonstrated for an ASTRID-like core within the ESNII+ framework studies performed by nine European institutions. In order to shift the paradigm towards best-estimate plus uncertainties, the nuclear data sensitivity analysis and uncertainty propagation on reactivity coefficients has to be carried out. The goal of this work is to assess the impact of nuclear data uncertainties on sodium voiding reactivity feedback coefficients in order to get a more complete picture of the actual safety margins of the ASTRID low void-core design. The nuclear data sensitivity analysis is performed in parallel using SCALE TSUNAMI-3D and the newly developed GPT SERPENT 2 module. A comparison is carried out between the two methodologies. Uncertainty on the sodium reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety margins conclusions are drawn.