Articles
Performance and safety evaluation of a <10 wt% 235U enriched small lead-cooled fast reactor
Annals of Nuclear Energy In Press (2024) 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.