Articles
Assessing the near-surface diffusion of Xe and Kr in Zirconia by time-of-flight elastic recoil detection analysis
Nuclear Instruments and Methods in Physics Research B 566 (2025) 1657736
N. Wikström, M. Giamouridou, E. Charatsidou, P. Olsson, J. Oscarsson, D. Primetzhofer, R.J.W. Frost
2025-06-06
Abstract
The diffusion of two volatile fission products, xenon (Xe) and krypton (Kr), in zirconia (ZrO2) is investigated. Samples of Yttria (Y2O3)-stabilised tetragonal ZrO2 were implanted with either Xe or Kr, at 300 keV, with a fluence of 10¹⁷ at./cm2, and subsequently analysed with time-of-flight elastic recoil detection analysis (ToF-ERDA) to obtain elemental composition depth profiles. Samples were then annealed at 1200 ◦C for 9 h, and the effect of the annealing was assessed by ToF-ERDA measurements. From these measurements, first-order approximations of diffusion coefficients for Xe and Kr in ZrO2 were derived, using a model based on Fick’s second law, these being (1.36 ± 0.87) × 10⁻¹⁹ m²/s and (2.94 ± 1.96) × 10⁻¹⁹ m²/s at 1200 ◦C for Kr and Xe respectively. It was shown that ToF-ERDA can provide data to analyse the diffusion of elements in solid sample matrices and that a model based on Fick’s Law can predict the diffusion of the implanted ions.
Impact of different TRU compositions on system response during an unprotected station blackout in small lead-cooled reactors
Annals of Nuclear Energy 222 (2025) 111586
Fredrik Dehlin, Janne Wallenius
2025-06-05
Abstract
The dynamic response to an Unprotected Station Blackout (USBO) has been evaluated for a small, lead-cooled reactor when fuelled with two different actinide compositions: one sourced from spent light water reactor (LWR) fuel and the other from UN fuel discharged from a small LFR. We demonstrate that a reduction in the delayed neutron fraction, primarily due to the addition of americium, leads to lower peak temperatures during phase one of the USBO. This reduction could help with ensuring cladding integrity despite an increased internal gas pressure resulting from helium production during the decay of 242Cm. It is also shown that the coolant volume required to buffer decay heat until vessel air cooling becomes effective must be increased to ensure the integrity of the fuel cladding. We conclude by demonstrating that (U,Pu)N fuel, with negligible 241Pu content, offers the best properties to ensure cladding integrity during the USBO.
A semi-analytical method for modelling station blackout transients in liquid metal-cooled reactors
Annals of Nuclear Energy 219 (2025) 111414
Janne Wallenius, Fredrik Dehlin
2025-04-13
Abstract
A semi-analytical method for modelling station blackout performance in liquid metal reactors is developed, permitting to identify key factors determining peak temperatures during the transient, and hence to design associated passive safety systems. It is shown that integrity of the fuel cladding during this transient can be ensured by adequate dimensioning of coolant channels, the primary system and the vessel air cooling circuit. These dimensions are determined using algebraic equations and postulated values for a minimum/maximum permissible Reynolds number, dimensionless parameters for the fuel cladding tube geometry and heat sink elevation, a guard vessel height, the nominal core power, permitted temperature gradients in the vessel air cooling system and the air cooling system chimney height. The model suggests that the required coolant volume is a rapidly growing function of core power, and that this volume needs to be 40% larger in a sodium-cooled reactor than in a lead-cooled reactor.
Effect of temperature and atmosphere on fretting wear of self-mated Fe-10Cr-4Al alloy for nuclear power application
Wear 564–565 (2025) 205704
Daria Kolbas, Leonardo Pelcastre, Braham Prakash, Jens Hardell
2025-03-15
Abstract
Over the years, fretting has been a leading cause of fuel failures for water cooled nuclear reactors. Implementation of liquid lead coolant increases the need for fretting and corrosion-resistant materials. Fe–10Cr–4Al alloy has shown good resistance to thermal aging, oxidation, liquid metal corrosion, and embrittlement in liquid lead environment. The performance of Fe-10Cr-4Al under fretting wear conditions has not been studied previously. This work aims to investigate the fretting wear and friction behaviour of self-mated Fe-10Cr-4Al alloy at varying temperatures, in the presence of a reducing gas environment and liquid lead. The results show that an increase in test temperature from RT to 550 °C decreases the coefficient of friction and promotes the formation of Cr and Fe oxide third body layer. At high temperature in reducing gas atmosphere, the initial plastic deformation and severe adhesion leads to early seizure. Presence of liquid lead has a lubricating effect and allows a third body layer to form on worn surfaces. A detailed description of the initial wear mechanisms of Fe-10Cr-4Al alloy and friction levels under various conditions provides a base for further investigation of this alloy for long-duration fretting wear.
A modified two-layer scalar diffusivity description for high Schmidt and Prandtl turbulent boundary layers
Physics of Fluids 37 (2025) 025219
Kin Wing Wong, Ignas Mickus, Dmitry Grishchenko, Pavel Kudinov
2025-02-26
Abstract
In engineering systems operating under high Schmidt (Sc) or Prandtl (Pr) number flow conditions, the demand for near-wall mesh refinement increases significantly, underscoring the need for cost-effective modeling approaches that avoid additional computational overhead. Existing models, which are predominantly designed for low-Sc flows, overlook temporal filtering effects, resulting in inaccuracies in theoretical description and mass transfer predictions. This paper addresses the impact of high Sc or Pr by refining the single-layer scalar diffusivity model. It introduces a switch between scalar filtering and eddy viscosity-dominated regions, leveraging two parameters: Sc, accounting for temporal filtering effects, and Re, addressing variations in Reynolds number. In addition, we adopted a complementary outer layer term to model the upwarding trend in low frictional Reynolds number condition. Using the two-layer model with unity Sc and/or Pr, a close agreement with the von-Kármán constant in the velocity boundary layer was observed. The modified model demonstrated strong agreement with scalar profiles across a broad range of Sc and friction Reynolds numbers (Reτ) in direct numerical simulation and large eddy simulation data, demonstrating its accuracy at low Reτ and predictive performance at high Reτ. The two-layer model improves the prediction of turbulent mass transfer, providing better alignment with high Sc engineering correlations than existing wall model approach. This study provides valuable insight for modeling the mass and heat transfer processes under high Sc or Pr conditions.
Influence of liquid lead and lead-bismuth eutectic on three alumina forming austenitic (AFA) steels through slow strain rate testing
Journal of Nuclear Materials 603 (2025) 155415
Christopher Petersson, Peter Szakalos, Rachel Pettersson, Mats Lundberg
2024-10-01
Abstract
Liquid metal embrittlement (LME) in three newly developed alumina-forming austenitic (AFA) alloys, two 50 kg batches and one 5-ton heat, was studied in the temperature range 350–600 °C in liquid Pb and 140–600 °C in LBE using slow strain rate testing (SSRT) in a low-oxygen environment. No significant decrease in the engineering strain was observed in either environment. However, the presence of secondary cracks along the length of the specimen and brittle intergranular areas on the fracture surfaces indicates that the AFA alloys do show a minor degree of embrittlement above 570 °C. This appears to be related to grain boundary wetting by Pb/LBE. At temperatures below 570 °C, this wetting effect does not seem to be strong enough to induce LME in the alloys, and their ability to form a sufficiently protective oxide means that they remain unaffected by LME. The results indicate that the AFA alloy group can perform sufficiently well in liquid Pb/LBE environments, and long-term testing should be carried out to determine their viability as candidate materials for use in Pb- and LBE-based cooling systems.
Performance and safety evaluation of a <10 wt% 235U enriched small lead-cooled fast reactor
Annals of Nuclear Energy 212 (2025) 110861
Fredrik Dehlin, Eloi Pallarès Abril, Janne Wallenius
2024-09-12
Abstract
We present the conceptual core design of a small lead-cooled fast reactor, for which a critical configuration has been achieved with a uranium enrichment of 9.9 wt%. This is a novelty for fast-neutron reactors without incorporating mixed uranium/plutonium fuel. It is shown how a reduction in uranium enrichment by two percentage points from a previously designed small lead-cooled reactor leads to an increase in conversion ratio of 20% and a significantly larger reactivity swing. The lowered enrichment gives a stronger Doppler feedback, which leads to lower temperatures during an overpower transient, despite remaining feedback coefficients being less negative. The new reactor geometry is presented along with a detailed neutronic characterisation, where whole-core reactivity feedback coefficients are derived, and depletion calculations are performed. Thereafter, we use the safety analysis code SAS4A/SASSYS-1 to demonstrate that the proposed design remains safe during enveloping unprotected transients, corresponding to Beyond Design Basis Accidents. We show how the reactor has a >2000°C margin to fuel melting during an Unprotected Overpower transient and that thermally induced creep rupture of the fuel cladding tubes is a non-issue despite conservatively assuming 100% fission gas release.
Role of various influencing parameters on high temperature fretting behaviour of different tribopairs in liquid lead
Nuclear Materials and Energy 40 (2024) 101699
Daria Kolbas, Leonardo Pelcastre, Braham Prakash, Jens Hardell
2024-06-27
Abstract
The increasing interest in liquid metal cooled nuclear reactors provides technical and scientific challenges such as the understanding, prevention, and prediction of the degradation of materials in liquid lead. Critical components include the fuel rods, heat exchanger tubes, and pump impellers. These functional elements are exposed to mechanical loading (up to 40 MPa), high temperatures (450–550 °C), and fluid-induced vibrations (up to 25 Hz). Under such conditions, fretting wear occurs between e.g., the spacer wire and the outer surface of the fuel or heat exchanger tubes. This work is aimed to establish a laboratory-scale fretting wear test setup and develop test methodology to enable systematic material characterisation in liquid metal environments. The results obtained by using the described methodology indicate that adhesive wear is the dominant degradation mechanism, and 316L stainless steel shows a higher coefficient of friction but a lower wear volume/tribolayer volume compared to 100Cr6 bearing steel. These results are in agreement with those reported in open literature and demonstrates the suitability of the presented method for conducting fretting tests and analysis for various materials and contact configurations in liquid lead environment.
First-principles predictions of structural and magnetic phase stability in irradiated α-Fe
Materials Research Letters 12 (2024) 477
Ebrahim Mansouri, Pär Olsson
2024-05-10
Abstract
We here use density functional theory and the creation-relaxation algorithm to investigate the appearance of polymorphism in α-Fe, driven by irradiation-induced microstructural changes. Local constriction leads to magnetic instability and provides excess energy required for structural phase transformation. Under extreme conditions, α-Fe undergoes local transformations into icosahedral C15 Laves phase with highly close-packed stacking and internal short-range ferromagnetic ordering, antiparallel to the bulk magnetisation. Analysing local magnetic moments and atomic volumes, in conjunction with the magneto-volume relations of different Fe structures, suggests two other alternatives for local phase transformation under irradiation conditions: the double-layer antiferromagnetic γ-Fe and non-magnetic ϵ-Fe.
Temperature-dependent thermal conductivity and fuel performance of UN-UO2 and UN-X-UO2 (X=Mo, W) composite nuclear fuels by finite element modeling
Journal of Materiomics 10 (2024) 937
Faris Sweidan, Diogo Ribeiro Costa, Huan Liu, Pär Olsson
2024-03-14
Abstract
The temperature-dependent effective thermal conductivity of UN-X-UO2 (X = Mo, W) nuclear fuel composite was estimated. Following the experimental design, the thermal conductivity was calculated using Finite Element Modeling (FEM), and compared with analytical models for 10%, 30%, 50%, and 70% (in mass) uncoated/coated UN microspheres in a UO2 matrix. The FEM results show an increase in the fuel thermal conductivity as the mass fraction of the UN microspheres increases – from 1.2 to 4.6 times the UO2 reference at 2000 K. The results from analytical models agree with the thermal conductivity estimated by FEM. The results also show that Mo and W coatings have similar thermal behaviors, and the coating thickness influences the thermal conductivity of the composite. At higher weight fractions, the thermal conductivity of the fuel composite at room temperature is substantially influenced by the high thermal conductivity coatings approaching that of UN. Thereafter, the thermal conductivity from FEM was used in the fuel thermal performance evaluation during LWR normal operation to calculate the maximum centerline temperature. The results show a significant decrease in the fuel maximum centerline temperature ranging from −94 K for 10% UN to −414 K for 70% (in mass) UN compared to UO2 under the same operating conditions.
Proton irradiation-induced cracking and microstructural defects in UN and (U,Zr)N composite fuels
Journal of Materiomics 10 (2024) 906
Elina Charatsidou, Maria Giamouridou, Andrea Fazi, Gyula Nagy, Diogo Ribeiro Costa, Sarmad Naim Katea, Mikael Jolkkonen, Gunnar Westin, Mattias Thuvander, Daniel Primetzhofer, Pär Olsson
2024-03-08
Abstract
Proton irradiation with a primary ion energy of 2 MeV was used to simulate radiation damage in UN and (U,Zr)N fuel pellets. The pellets, nominally at room temperature, were irradiated to peak levels of 0.1, 1, 10 dpa and 100 dpa resulting in a peak hydrogen concentration of at most 1 at. %. Microstructure and mechanical properties of the samples were investigated and compared before and after irradiation. The irradiation induced an increase in hardness, whereas a decrease in Young’s modulus was observed for both samples. Microstructural characterization revealed irradiation-induced cracking, initiated in the bulk of the material, where the peak damage was deposited, propagating towards the surface. Additionally, transmission electron microscopy (TEM) was used to study irradiation defects. Dislocation loops and fringes were identified and observed to increase in density with increasing dose levels. The high density of irradiation defects is proposed as the main cause of swelling and consequent sample cracking, leading simultaneously to increased hardening and a decrease in Young’s modulus.
Hydrodynamic design of the Separate Effect test facility for Flow-Accelerated Corrosion and Erosion (SEFACE) studies in liquid lead
Nuclear Engineering and Design 417 (2024) 112852
Kin Wing Wong, Ignas Mickus, Nathaniel Torkelson, Sumathi Vasudevan, Haipeng Li, Dmitry Grishchenko, Pavel Kudinov
2024-02-01
Abstract
Flow-accelerated corrosion and erosion (FACE) phenomena can be crucial for performance of structural elements in heavy liquid metal (HLM) cooled reactor systems. Existing experimental observations indicate that turbulent flow characteristic can affect FACE, but there is no quantitative data that can be used for model development and validation. Main recirculation pump impellers, which operate at high relative velocities and rotational flow conditions can be especially vulnerable to FACE. For comparison, the core internals operate at lower velocities and in axial flow conditions, but at higher temperatures and neutron fluence. Hence, systematic experimental data is needed to improve our knowledge on FACE phenomena. The Separate Effect Test Facility for Flow-Accelerated Corrosion and Erosion (SEFACE) is designed to obtain such experimental data including high relative velocities (up 20 ms−1) and high temperatures (400 to 550 °C) of liquid lead. This article focuses on the hydrodynamic design of SEFACE. The aim of the design is to achieve well defined flow conditions for experiments and ensure safe operation of the facility. First, we examine three design concepts (i.e., forced convection loop, rotating cylinder, and rotating disk) and motivate the choice of the rotating disk approach for SEFACE. Second, we discuss different design options, i.e., a confined rotor–stator test chamber and the unconfined rotating disk configuration. We used Reynolds-Averaged Navier Stokes (RANS) calculations to identify and solve the issues stemming from the high rotational speed. These include, for instance, lead free surface deformation, radial pressure buildup, and axial bending forces due to asymmetric test chamber. The CFD-derived torque and power predictions in rotor–stator and rotating disk systems are verified with selected empirical turbulent friction factor correlations or/and DNS calculations. We demonstrate that the developed hydrodynamic design of SEFACE solves identified issues and enables obtaining experimental data under well-defined flow conditions. The findings are deemed to also be applicable to the design of rotating disk-type FACE installations for other liquid mediums.
Modeling of irradiation-induced microstructure evolution in Fe: Impact of Frenkel pair distribution
Computational Materials Science 236 (2024) 112852
Ebrahim Mansouri, Pär Olsson
2024-02-10
Abstract
This study investigated the irradiation-induced microstructure evolution in Fe, employing the Creation-Relaxation Algorithm and different interatomic potentials. The influence of self-interstitial atoms (SIAs), which were either locally or uniformly being distributed during the creation of the Frenkel pairs, was investigated on the evolving microstructure. The spatially localized distribution of SIAs, mimicking the low-energy transfer irradiation conditions, moderated the microstructure development, compared to uniform distribution of SIAs, delaying the nucleation of dislocation for higher irradiation doses. Introducing multiple Frenkel pairs facilitated a cumulative irradiation dose of 5 dpa in large supercells. In small supercells, the accumulation of SIAs led to the formation of an artificially stabilized self-interacting planar interstitial cluster, suggesting a minimum cell dimension of 10 nm for an accurate modeling of microstructure evolution when the development of the dislocation network is of interest. The formation and evolution of the C15 Laves phase structure were explored. The evolving C15 structure developed larger clusters with uniformly distributed SIAs, and their sizes depended on the interatomic potential employed. Finally, a comparison with experimental measurements demonstrated that the density and the average size of interstitial dislocation loops aligned well with those observed in experimentally irradiated ultra-high purity Fe at low and room temperatures.
Microstructure and magnetization evolution in bcc iron via direct first-principles predictions of radiation effects
Physical Review Materials 7 (2023) 123604
Ebrahim Mansouri, Pär Olsson
2023-12-22
Abstract
We here model the radiation-induced microstructure evolution in bcc iron using the creation-relaxation algorithm directly driven by density functional theory calculations and compare to interatomic potential simulations. The method is in essence a relatively simplified model without thermally driven diffusion. The microstructure evolution in this model is driven mostly by the stress field introduced by sequential direct damage insertions. The defect populations and the corresponding defect cluster characteristics have been analyzed as a function of irradiation dose. Different distribution functions have been investigated for the self-interstitial atom implantation, to make model predictions as close as possible to actual irradiation conditions under which defects are produced. The stability and magnetic characteristics of the defect structures that are formed are studied. Our first-principles simulations revealed that the C15 structure can be dynamically formed during the irradiation of the material and that the constituent atoms align antiferromagnetically to the lattice. For doses on the order of a fraction of 1 displacement per atom, the model material Fe experiences mechanical changes caused by continuous irradiation and approaches a saturation state of about 2% swelling. The radiation-induced change in the local magnetic moments as well as the charge density differences for some isolated and clustered defects have been investigated. The results revealed a net reduction in total magnetization per atom and a tendency for interstitial sites to have a spin polarization opposing the intrinsic atomic site spins when the coordination number was increased compared to that of the initial lattice structure. It is suggested that radiation-induced damage could be traced using nondestructive measures of bulk magnetization changes.
Negative effect of bismuth in lead on liquid metal embrittlement of a ferritic steel
Journal of Nuclear Materials 588 (2024) 154829
Christopher Petersson, Peter Szakalos, Rachel Pettersson, Daniel Dietrich Stein
2023-11-22
Abstract
Liquid metal embrittlement (LME) of a Fe-10Cr-4Al ferritic steel was studied at 375 °C in liquid Pb-Bi alloys. Slow strain rate testing (SSRT) in low oxygen conditions was used to evaluate the ductility as a function of Bi concentration. It was found that susceptibility to LME increased strongly with the Bi concentration. The steel showed a reduction in its total strain to failure, which started at 3–5 wt.% Bi. The alloying elements (Fe, Cr, and Al) have a higher solubility in Bi than pure lead (Pb), so they are expected to dissolve more readily when Bi is added to the Pb. This is believed to be part of the explanation for the observed increase of LME. Lead with up to 3 wt.% Bi induced no LME in the studied corrosion resistance FeCrAl alloy.
WC-Ni cemented carbides prepared from Ni nano-dot coated powders
International Journal of Refractory Metals and Hard Materials 117 (2023) 106375
Paul H. Gruber, Sarmad Naim Katea, Gunnar Westin, Farid Akhtar
2023-08-24
Abstract
This study presents a novel approach for the synthesis of WC-Ni cemented carbides with enhanced mechanical properties. A low-cost solution-based route was used to coat WC powders with well-distributed metallic nickel dots measuring between 17 nm and 39 nm in diameter. Varying compositions with loadings of 2, 6, and 14 vol% Ni were consolidated using spark plasma sintering (SPS) at 1350 °C under 50 MPa of uniaxial pressure giving relative densities of 99 ± 1 %. The sintered WC-Ni cemented carbides had an even distribution of the Ni binder phase in all compositions, with retained ultrafine WC grain sizes of 0.5 ± 0.1 μm from the starting powder. The enhanced sinterability of the coated powders allowed for consolidation to near theoretical densities, with a binder content as low as 2 vol%. This is attributed to the uniform distribution of nickel and an extensive Ni-WC interface existing prior to sintering. The small size of the Ni dots likely also contributed to the solid-state sintering starting temperatures of as low as 800 °C. The mechanical performance of the resulting cemented carbides was evaluated by measuring the hardness at temperatures between 20 °C and 700 °C and estimating toughness at room temperature using Vickers indentations. These results showed that the mechanical properties of the WC-Ni cemented carbides synthesised by our method were comparable to conventionally prepared WC-Co cemented carbides with similar grain sizes and binder contents and superior to conventionally prepared WC-Ni cemented carbides. In particular, the 2 vol% Ni composition had excellent hardness at room temperature of up to 2210HV10, while still having an indentation fracture toughness of 7 MPa·m0.5. Therefore, WC-Ni cemented carbides processed by this novel approach are a promising alternative to conventional WC-Co cemented carbides for a wide range of applications.
Activation analysis of the lead coolant in SUNRISE-LFR
Nuclear Engineering and Design 414 (2023) 112503
Fredrik Dehlin, Janne Wallenius
2023-08-04
Abstract
A lumped, zero-dimensional, mass transport model is combined with a depletion matrix solver to study the influence of coolant circulation on radionuclide build-up in a small lead-cooled fast reactor. It is shown that the addition of coolant circulation results in a lower activity for a minority of studied nuclides, and it is thus recommended to consider stagnant coolant when licensing a reactor. Activation analysis of three different lead qualities potentially used in SUNRISE-LFR is performed, and the result shows that a low silver content is desirable to simplify maintenance and decommissioning.
Slow strain rate testing of Fe-10Cr-4Al ferritic steel in liquid lead and lead-bismuth eutectic
Nuclear Materials and Energy 34 (2023) 101403
Christopher Petersson, Peter Szakalos, Daniel Stein
2023-03-20
Abstract
The susceptibility of Fe-10Cr-4Al steel to liquid metal embrittlement (LME) in low oxygen liquid lead and lead–bismuth eutectic (LBE) environments has been investigated using a newly developed slow strain rate testing (SSRT) technique that can be employed at elevated temperatures. This study showed that the Fe-10Cr-4Al steel suffered embrittlement when exposed to LBE over a wide temperature range. The embrittlement, here measured as a reduction in fracture strain, was observed at the melting temperature of LBE and reached a maximum at approximately 375 °C. At temperatures above 425 °C, the material's ductility regained its original levels. The exposures in liquid lead showed no indications of embrittlement, but a ductile behavior over the entire temperature range studied (340–480 °C).
Potential accident tolerant fuel candidate: Investigation of physical properties of the ternary phase U2CrN3
Journal of Nuclear Materials 568 (2022) 153851
Yulia Mishchenko, Sobhan Patnaik, Elina Charatsidou, Janne Wallenius, Denise Adorno Lopes
2022-09-25
Abstract
In the present study, physical properties of the ternary phase U2CrN3 are evaluated experimentally and by modeling methods. High density pellets containing the ternary phase were prepared by spark plasma sintering (SPS). The microstructural and crystallographic analyses of the composite pellets were performed using scanning electron microscopy (SEM), standardised energy dispersive spectroscopy (EDS) and electron backscatter diffraction (EBSD). Evaluation of the mechanical properties was performed by nanoindentation test. The impact of temperature on lattice properties was evaluated using high temperature X-ray diffraction (XRD) coupled with modeling. Progressive change in the lattice parameters was obtained from room temperature (RT) to 673 K, and the result was used to calculate average linear thermal expansion coefficients, as well as an input for the density functional theory (DFT) modeling to reassess the degradation of the mechanical properties. The ab-initio calculation provides an initial assessment of electronic configuration of this ternary phase in a direct comparison with one of UN phase. For this goal, modeling was also employed to evaluate point defect formation energies and electronic charge distribution in the ternary phase. Results indicate that the U2CrN3 phase has similar mechanical properties to UN (Young's, bulk, shear moduli, hardness). No preferential crystallographic orientation was observed in the composite pellet. However, charge electron density distribution highlights the significant directionality of chemical bonds, which is in agreement with the anisotropy and non-linear behaviour of the obtained thermal expansion ((aa) = 9.12×10⁻⁶/K, (ab) = 5.81×10⁻⁶/K and (ac) = 6.08×10⁻⁶/K). As a consequence, uranium was found to be more strongly bound in the ternary structure which may delay diffusion and vacancy formation, promising an acceptable performance as nuclear fuel.
An improved correlation for gas release from nitride fuels
Journal of Nuclear Materials 558 (2022) 153402
Janne Wallenius
2022-01-31
Abstract
An improved correlation for gas release from nitride fuels is elaborated. Introducing empirical activation energies for migration of fission gases in presence of solid fission products and oxide impurities, it becomes possible to better reproduce existing experimental data sets for gas release in sodium and helium bonded rods. The suggested approach may assist in resolving the previously poorly understood dispersion in measured gas release for identical irradiation conditions.
An analytic approach to the design of passively safe lead-cooled reactors
Annals of Nuclear Energy 169 (2022) 108971
Fredrik Dehlin, Janne Wallenius, Sara Bortot
2022-01-24
Abstract
A methodology to assist the design of liquid metal reactors, passively cooled by natural circulation during off-normal conditions, is derived from first principle physics. Based on this methodology, a preliminary design of a small LFR is accomplished and presented with accompanying neutronic and reactor dynamic characterizations. The benefit of using this methodology for reactor design compared to other available methods is discussed.