Imperial College London

DrMarkWenman

Faculty of EngineeringDepartment of Materials

Reader in Nuclear Materials
 
 
 
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Contact

 

+44 (0)20 7594 6763m.wenman

 
 
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Location

 

B301aRoyal School of MinesSouth Kensington Campus

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Summary

 

Publications

Publication Type
Year
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89 results found

Battistini A, Haynes TA, Shepherd D, Wenman MRet al., 2023, Residual stresses in as-manufactured TRISO Coated Particle Fuel (CPF), Journal of Nuclear Materials, Vol: 586, Pages: 154659-154659, ISSN: 0022-3115

Journal article

Nicholls O, Frost D, Tuli V, Smutna J, Wenman MR, Burr PAet al., 2023, Transferability of Zr-Zr interatomic potentials, Journal of Nuclear Materials, Vol: 584, ISSN: 0022-3115

Tens of Zr inter-atomic potentials (force fields) have been developed to enable atomic-scale simulations of Zr alloys. These can provide critical insight in the in-reactor behaviour of nuclear fuel cladding and structural components exposed, but the results are strongly sensitive to the choice of potential. We provide a comprehensive comparison of 13 popular Zr potentials, and assess their ability to reproduce key physical, mechanical, structural and thermodynamic properties of Zr. We assess the lattice parameters, thermal expansion, melting point, volume-energy response, allotropic phase stability, elastic properties, and point defect energies, and compare them to experimental and ab-initio values. No potential was found to outperform all others on all aspects, but for every metric considered here, at least one potential was found to provide reliable results. Older embedded-atom method (EAM) potentials tend to excel in 2-3 metrics each, but at the cost of poorer transferability. The two highest-performing potentials overall, with complementary strengths and weaknesses, were the 2021 angular-dependent potential of Smirnova and Starikov (Comp. Mater. Sci. 197, 110581) and the 2019 embedded-atom method potential of Wimmer et al (J. Nucl. Mater. 532, 152055). All potentials trained through machine learning algorithms proved to have lower overall accuracy, and less transferability, than simpler and computationally faster potentials available. Point defect structures and energies is where the greatest divergence and least accuracy is observed. We created maps that will help modellers select the most suitable potential for a specific application, and which may help identify areas of improvement in future potentials.

Journal article

Haynes TA, Battistini A, Leide AJ, Liu D, Jones L, Shepherd D, Wenman MRet al., 2023, Peridynamic modelling of cracking in TRISO particles for high temperature reactors, Journal of Nuclear Materials, Vol: 576, Pages: 1-12, ISSN: 0022-3115

A linear-elastic computer simulation (model) for a single particle of TRISO fuel has been built using a bond-based peridynamic technique implemented in the finite element code ‘Abaqus’. The model is able to consider the elastic and thermal strains in each layer of the particle and to simulate potential fracture both within and between layers. The 2D cylindrical model makes use of a plane stress approximation perpendicular to the plane modelled. The choice of plane stress was made by comparison of 2D and 3D finite element models. During an idealised ramp to normal operating power for a kernel of 0.267 W and a bulk fuel temperature of 1305 K, cracks initiate in the buffer near to the kernel-buffer interface and propagate towards the buffer-iPyC coating interface, but do not penetrate the iPyC and containment of the fission products is maintained. In extreme accident conditions, at around 600% (1.60 W) power during a power ramp at 100% power (0.267 W) per second, cracks were predicted to form on the kernel side of the kernel-buffer interface, opposite existing cracks in the buffer. These were predicted to then only grow further with further increases in power. The SiC coating was predicted to subsequently fail at a power of 940% (2.51 W), with cracks formed rapidly at the iPyC-SiC interface and propagating in both directions. These would overcome the containment to fission gas release offered by the SiC ‘pressure vessel’. The extremely high power at which failure was predicted indicates the potential safety benefits of the proposed high temperature reactor design based on TRISO fuel.

Journal article

Liu J, Gasparrini C, White JT, Johnson K, Lopes DA, Peterson VK, Studer A, Griffiths GJ, Lumpkin GR, Wenman MR, Burr PA, Sooby ES, Obbard EGet al., 2023, Thermal expansion and steam oxidation of uranium mononitride analysed via in situ neutron diffraction, JOURNAL OF NUCLEAR MATERIALS, Vol: 575, ISSN: 0022-3115

Journal article

Smutna J, Wenman MR, Horsfield AP, Burr PAet al., 2023, The bonding of H in Zr under strain, Journal of Nuclear Materials, Vol: 573, Pages: 1-12, ISSN: 0022-3115

Accurate computer simulation is important for understanding the role of irradiation-induced defects in zirconium alloys found in nuclear reactors. Of particular interest is the distribution and trapping of hydrogen, and the formation of zirconium hydride. These simulations require an accurate representation of Zr-H bonding in order to predict the behaviour of H around atomic-scale defects, dislocation lines, and dislocation loops. Here we explore the bonding of H in Zr under strain, how well it is represented by state-of-the-art Embedded Atom Method (EAM) potentials, and what physics is needed for an accurate representation in a Linear Combination of Atomic Orbitals (LCAO) Density Functional Theory (DFT) framework. For H in dilute solution under hydrostatic strain in the range -10% to +10%, solution energies and Zr-H bond lengths computed using EAM potentials are found to be in poor agreement with plane-wave DFT results. We note that the bond lengths are in a poor agreement even in equilibrium. LCAO basis sets are used to explore the importance of electron distribution around H atoms, and the transfer of electrons between H and Zr. The electron distribution around H atoms is found to be important to the explanation of the difference between octahedral and tetrahedral interstitial sites for H, with H in a tetrahedral site having very similar bonding to H in zirconium hydrides. The interatomic electron transfer has a smaller impact but is needed for maximum accuracy.

Journal article

Jones LD, Haynes TA, Rossiter G, Wenman MRet al., 2022, Application of Weibull fracture strength distributions to modelling crack initiation behaviour in nuclear fuel pellets using peridynamics, JOURNAL OF NUCLEAR MATERIALS, Vol: 572, ISSN: 0022-3115

Journal article

Reali L, Balint DS, Wenman MR, 2022, Discrete dislocation modelling of ? hydrides in Zr: towards an understanding of the importance of interfacial stresses for crack initiation, JOURNAL OF NUCLEAR MATERIALS, Vol: 572, ISSN: 0022-3115

Journal article

Reali L, Balint DS, Wenman MR, 2022, Dislocation modelling of the plastic relaxation and thermal ratchetting induced by zirconium hydride precipitation, Journal of the Mechanics and Physics of Solids, Vol: 167, Pages: 104988-104988, ISSN: 0022-5096

The precipitation of hydrides in zirconium alloys is accompanied by a significant and anisotropic volumetric expansion. Previous literature quantified the misfit both theoretically and experimentally, but these values differ greatly; the experimental values are consistently lower. One possibility is that the experimental measurements include the effect of dislocations generated by the hydride, which relax the transformation stresses. To test this hypothesis, it is important to determine the stress field of a hydride and its associated dislocations, combined. A simple planar dislocation model was developed of the hydride—dislocation ensemble in -Zr. By capturing details of the dislocation structures given in the literature, it is shown in this study that including the interfacial dislocations largely reconciles the predicted and experimental values. Discrete dislocation plasticity is then used to model the diffuse plastic relaxation associated with hydride formation. The effects of plastic relaxation on the equilibrium hydrogen profile, hence the implications for subsequent hydride precipitation, are discussed. In particular, precipitation–dissolution cycles were simulated to calculate the magnitude of the residual hydrostatic tension, which is argued to be the primary cause of the “memory effect” for the re-precipitation of both and hydrides.

Journal article

Wenman M, Grimes R, Tate J, Bluck M, Davies C, Eaton M, Lawrence J, Dini D, Appelbe Bet al., 2022, Parliament Committees. Delivering nuclear power: Written evidence from Imperial College London (NCL0026), Publisher: UK Parliment

Report

Podgurschi V, King DJM, Luo K, Wenman MRet al., 2022, Atomic scale simulation of the strain rate and temperature dependence of crack growth and stacking faults in zirconium, Computational Materials Science, Vol: 206, Pages: 1-12, ISSN: 0927-0256

Molecular dynamics simulations of single crystal zirconium fracture were performed to study thedeformation mechanisms active on the basal and prismatic planes. The effects of temperature (0 to300 K) and strain rate (108–1010 s−1) were investigated. Crack tip orientation was found to stronglyaffect the fracture behaviour. On the basal plane twinning ({11¯21}<1¯126>) and emission of <c +a> type dislocations that then dissociated into partial dislocations around pyramidal I2 stackingfaults were seen to occur during fracture. At higher strain rates (109 and 1010 s−1), twinningoccurred. The emission of edge dislocations ( 13<1¯210> type) was prevalent on the prismatic planeand were found to be strongly affected by temperature. At higher temperature (150 and 300 K), thedislocation density increased. The crack grew further at 150–300 K than at 0 K and the shieldingeffect of dislocations was limited due to their movement away from the crack tip. The addition ofiodine at basal I2, pyramidal I1 and I2 stacking faults was seen to decrease the energy of its formation whereas for the prismatic stacking fault it was found to increase it. The iodine also changed theorder of favourability of the stacking faults with basal I2 and pyramidal I1 stacking faults becomingmuch more favourable and prismatic going from most to least favourable.

Journal article

King DJM, Knowles AJ, Bowden D, Wenman MR, Capp S, Gorley M, Shimwell J, Packer L, Gilbert MR, Harte Aet al., 2022, High temperature zirconium alloys for fusion energy, Journal of Nuclear Materials, Vol: 559, Pages: 1-27, ISSN: 0022-3115

This review considers current Zr alloys and opportunities for advanced zirconium alloys to meet the demands of a structural material in fusion reactors. Zr based materials in the breeder blanket offer the potential to increase the tritium breeding ratio above that of Fe, Si and V based materials. Current commercial Zr alloys might be considered as a material in water-cooled breeder blanket designs, due to the similar operating temperature to fission power plants. For breeder blankets designed to operate at higher temperatures, current commercial Zr alloys will not meet the high temperature strength and thermal creep requirements. Hence, Zr alloys with an operational temperature capability beyond that of current commercial fission alloys have been reviewed, specifically: binary Zr alloy systems Zr-Al, Zr-Be, Zr-Cr, Zr-Nb Zr-Ti, Zr-Si, Zr-Sn, Zr-V and Zr-W; as well as higher order Zr alloys Zr-Mo-Ti, Zr-Nb-Ti, Zr-Ti-Al-V and Zr-Mo-Sn. It is concluded that, with further work, higher order Zr alloys could achieve the required high temperature strength, alongside ductility, while maintaining a low thermal neutron cross-section. However, there is limited data and uncertainty regarding the structural performance and microstructural stability of the majority of advanced Zr alloys for temperatures 500–700 °C, at which they would be expected to operate for helium- and liquid metal-cooled breeder blanket designs.

Journal article

Carter M, Gasparrini C, Douglas JO, Riddle N, Edwards L, Bagot PAJ, Hardie CD, Wenman MR, Moody MPet al., 2022, On the influence of microstructure on the neutron irradiation response of HIPed SA508 steel for nuclear applications, JOURNAL OF NUCLEAR MATERIALS, Vol: 559, ISSN: 0022-3115

Journal article

Podgurschi V, King DJM, Smutna J, Kermode JR, Wenman MRet al., 2022, Atomistic modelling of iodine-oxygen interactions in strained sub-oxides of zirconium, Journal of Nuclear Materials, Vol: 558, Pages: 1-10, ISSN: 0022-3115

In water reactors, iodine stress corrosion cracking is considered the cause of pellet-cladding interaction failures, but the mechanism and chemistry are debated and the protective effect of oxygen is not understood. Density functional theory calculations were used to investigate the interaction of iodine and oxygen with bulk and surface Zr under applied hydrostatic strain (2% to +3%) to simulate crack tip conditions in Zr to ZrO, using a variety of intermediate suboxides (ZrO, ZrO, ZrO and ZrO). The formation energy of an iodine octahedral interstitial in Zr was found to decrease with increasing hydrostatic strain, whilst the energy of an iodine substitutional defect was found to be relatively insensitive to strain. As the oxygen content increased, the formation energy of an iodine interstitial increased from 1.03 eV to 8.61 eV supporting the idea that oxygen has a protective effect. At the same time, a +3% tensile hydrostatic strain caused the iodine interstitial formation energy to decrease more in structures with higher oxygen content: 4.56 eV decrease in ZrO compared to 1.47 eV decrease for pure Zr. Comparison of the substitutional and interstitial energies of iodine, to the adsorption energy of iodine, in the presence of oxygen, shows the substitutional energy of iodine onto a Zr site is more favourable for all strains and even interstitial iodine is favourable between strains of +1-5%. Although substitutional defects are preferred to octahedral interstitial defects, in the ordered suboxides, a 3% tensile strain significantly narrows the energy gap and higher strains could cause interstitial defects to form.

Journal article

Stephens GF, Than YR, Neilson W, Evitts LJ, Wenman MR, Murphy ST, Grimes RW, Cole-Baker A, Ortner S, Gotham N, Rushton MJD, Lee WE, Middleburgh SCet al., 2021, The accommodation of lithium in bulk ZrO<sub>2</sub>, SOLID STATE IONICS, Vol: 373, ISSN: 0167-2738

Journal article

Liu Y, El Chamaa S, Wenman MR, Davies CM, Dunne FPEet al., 2021, Hydrogen concentration and hydrides in Zircaloy-4 during cyclic thermomechanical loading, Acta Materialia, Vol: 221, Pages: 1-16, ISSN: 1359-6454

Hydride formation in Zircaloy-4 under cyclic thermomechanical loading has been investigated using characterized notched beam samples in four-point beam testing, and microstructurally-representative crystal plasticity modelling of the beam tests which incorporates an atomistically-informed equilibrium-state model for hydrogen concentration. The model provided the locations within the microstructure of high hydrogen content, above that required for saturation, hence predicting the anticipated locations of hydride observations in the experiments. The strain rate sensitivity of this alloy over the temperature range considered led to considerable intragranular slip and corresponding stress redistribution, and cyclic strain ratcheting leading to high hydrostatic stresses and in turn hydrogen concentrations, which explains the locations of experimentally observed hydride formation. The interstitial hydrogen interaction energy as well as the intragranular geometrically necessary dislocation density were shown to be important in controlling the spatial distributions of observed hydrides.

Journal article

Reali L, Balint DS, Sutton A, Wenman Met al., 2021, Plastic relaxation and solute segregation to β-Nb second phase particles in Zr-Nb alloys: a discrete dislocation plasticity study, Journal of the Mechanics and Physics of Solids, Vol: 156, ISSN: 0022-5096

There is clear evidence in the literature that iron segregates to the interface of second phase particles (SPPs) in unirradiated Zr-Nb alloys, and that it does not do so in the presence of radiation damage. In this work, a discrete dislocation plasticity model is developed that takes into account the long-range stress field of the SPP interface. A simple analytical model is also outlined, providing an upper bound for estimating the amount of interstitial segregation. The model provides a possible mechanism to explain both the iron segregation to coherent SPPs and its subsequent loss after irradiation. Qualitatively, the model proved to be insensitive to variations of all geometrical and computational parameters, allowing for general conclusions to be drawn. The model suggests that the segregation originates from a tensile field of order 1 GPa induced by the dislocations generated during the plastic relaxation around the SPP. This leads to the six-fold increase in the iron concentration observed in experiments. In the model, the loss of SPP/matrix coherency after irradiation causes the dislocations to drift away from the interface, and the iron concentration is homogenised accordingly. The hydrogen concentration was also predicted and found to be about 50% higher than in the bulk zirconium matrix at room temperature. The computational framework is built to be fast, making possible a statistical analysis on over five hundred simulations for improved reliability of the predictions.Keywords: Discrete dislocation plasticity; Zr-Nb alloys; second phase particles; interfacial segregation.

Journal article

Patel M, Reali L, Sutton AP, Balint DS, Wenman MRet al., 2021, A fast efficient multi-scale approach to modelling the development of hydride microstructures in zirconium alloys, COMPUTATIONAL MATERIALS SCIENCE, Vol: 190, ISSN: 0927-0256

Journal article

Dong P, Vecchiato F, Yang Z, Hooper PA, Wenman MRet al., 2021, The effect of build direction and heat treatment on atmospheric stress corrosion cracking of laser powder bed fusion 316L austenitic stainless steel, ADDITIVE MANUFACTURING, Vol: 40, ISSN: 2214-8604

Journal article

Yang M, King DJM, Postugar I, Wen Y, Luan J, Kuhn B, Jiao Z, Wang C, Wenman MR, Liu Xet al., 2021, Precipitation behavior in G-phase strengthened ferritic stainless steels, ACTA MATERIALIA, Vol: 205, ISSN: 1359-6454

Journal article

Reali L, Wenman MR, Sutton AP, Balint DSet al., 2021, Plasticity of zirconium hydrides: a coupled edge and screw discrete dislocation model, JOURNAL OF THE MECHANICS AND PHYSICS OF SOLIDS, Vol: 147, ISSN: 0022-5096

Journal article

Fallah AS, Giannakeas IN, Mella R, Wenman MR, Safa Y, Bahai Het al., 2020, On the computational derivation of bond-based peridynamic stress tensor, Journal of Peridynamics and Nonlocal Modeling, Vol: 2, Pages: 352-378, ISSN: 2522-896X

The concept of ‘contact stress’, as introduced by Cauchy, is a special case of a nonlocal stress tensor. In this work, the nonlocal stress tensor is derived through implementation of the bond-based formulation of peridynamics that uses an idealised model of interaction between points as bonds. The method is sufficiently general and can be implemented to study stress states in problems containing stress concentration, singularity, or discontinuities. Two case studies are presented, to study stress concentration around a circular hole in a square plate and conventionally singular stress fields in the vicinity of a sharp crack tip. The peridynamic stress tensor is compared with finite element approximations and available analytical solutions. It is shown that peridynamics is capable of capturing both shear and direct stresses and the results obtained correlate well with those obtained using analytical solutions and finite element approximations. A built-in MATLAB code is developed and used to construct a 2D peridynamic grid and subsequently approximate the solution of the peridynamic equation of motion. The stress tensor is then obtained using the tensorial product of bond force projections for bonds that geometrically pass through the point. To evaluate the accuracy of the predicted stresses near a crack tip, the J-integral value is computed using both a direct contour approximation and the equivalent domain integral method. In the formulation of the contour approximation, bond forces are used directly while the proposed peridynamic stress tensor is used for the domain method. The J-integral values computed are compared with those obtained by the commercial finite element package Abaqus 2018. The comparison provides an indication on the accurate prediction of the state of stress near the crack tip.

Journal article

Whiting TM, King DJM, Wenman MR, 2020, On the formation and structure of Mn-Ni-Si Γ2 precipitates in steels, Journal of Nuclear Materials, Vol: 542, Pages: 1-7, ISSN: 0022-3115

The Γ2-phase is postulated to form from solute clusters in neutron irradiated low-alloy steels. Density functional theory calculations were used to identify the ground state Γ2 structures with variation in Si and Ni contents. It was found that favourability of the Γ2-phase was proportional to Si content, however Si-Si first nearest neighbours reduced favourability. The substitutional enthalpies of Al, Cr, Cu, Fe, Ge, Hf, Mo, Nb, P, Ta, Ti, V, Zr, and vacancies into the ground state Γ2-phase structure from a ferrite matrix were calculated. It was found that Zr, Hf, Nb, Cu, Ti and Ta favourably substitute onto the Mn sites, Cu and P onto the Si sites and Cu onto the Ni sites and vacancies will substitute favourably onto all atomic sites. Finally, it is shown that, for ≤20 at% Fe concentrations, the Γ2-phase becomes more favourable than the bcc structure therefore it is plausible that the Γ2-phase could be thermodynamically stabilised provided these conditions are met.

Journal article

Vecchiato FL, de Winton H, Hooper PA, Wenman MRet al., 2020, Melt pool microstructure and morphology from single exposures in laser powder bed fusion of 316L stainless steel, Additive Manufacturing, Vol: 36, Pages: 101401-101401, ISSN: 2214-8604

Journal article

Fogarty R, Smutna J, Wenman M, Horsfield Aet al., 2020, Beyond two-center tight binding: Models for Mg and Zr, Physical Review Materials, Vol: 4, ISSN: 2475-9953

We describe a systematic approach to building ab initio tight-binding models and apply this to hexagonal metals Mg and Zr. Our models contain three approximations to plane-wave density functional theory (DFT): (i) we use a small basis set, (ii) we approximate self-consistency, and (iii) we approximate many-center exchange and correlation effects. We test a range of approximations for many-center exchange-correlation and self-consistency to gauge the accuracy of each in isolation. This systematic approach also allows us to combine multiple approximations in the optimal manner for our final tight-binding models. Furthermore, the breakdown of errors into those from individual approximations is expected to be a useful guide for which approximations to include in other tight-binding models. We attempt to correct any remaining errors in our models by fitting a pair potential. Our final tight-binding model for Mg shows excellent agreement with plane-wave results for a wide range of properties (e.g., errors below 10% for self-interstitial formation energies and below 3% for equilibrium volumes) and is expected to be highly transferable due to the minimal amount of fitting. Calculations with our Zr model also show good agreement with plane-wave results (e.g., errors below 2% for equilibrium volumes) except for properties where self-consistency is important, such as self-interstitial formation energies. However, for these properties we are able to generate a tight-binding model which shows excellent agreement with non-self-consistent DFT with a small basis set (i.e., many-center effects are captured accurately by our approximations). As we understand the source of remaining errors in our Zr model we are able to outline the methods required to build upon it to describe the remaining properties with greater accuracy.

Journal article

Haynes TA, Shepherd D, Wenman MR, 2020, Preliminary modelling of crack nucleation and propagation in SiC/SiC accident-tolerant fuel during routine operational transients using peridynamics, Journal of Nuclear Materials, Vol: 540, ISSN: 0022-3115

Silicon carbide fibre in silicon carbide matrix composites (SiC/SiC) are a promising cladding for use in accident tolerant fuels (ATF) in current light water reactor (LWR) designs. However, as they are a radically different material from current metal clads, current thermomechanical simulation methods struggle to accurately predict their behaviour, especially regarding the potential development of cracks. Thus, a new peridynamic model for SiC/SiC cladding has been developed in the Abaqus finite element code. The material model was isotropic and considers matrix cracking and fibre pull-out. The thermal expansion, swelling and the degradation of the thermal conductivity are modelled under typical LWR irradiation conditions. The swelling on the outer surface is predicted to be greater than the inner surface due to the lower irradiation temperature, causing a tensile stress on the inside of the cladding; tension being more challenging for a ceramic than a metal. This stress increases during the decrease in power at the start of a typical pressurised water reactor refuelling outage and causes microcracking of the matrix on the cladding inner surface. In models without fibres, cracks would propagate through the cladding. If fibres are modelled, matrix cracking will extend to a depth of around 20% through the cladding from the inner surface, which is unlikely to be an acceptable design. If an inner monolith of SiC is additionally modelled, cracking propagates through the monolith and acts as a stress raiser for matrix cracking in the composite, and therefore does not constitute a design improvement. If an outer SiC monolith is modelled, fibre pull-out strain on the inner surface of the cladding was increased by just under 70%. No cracks are predicted in an outer monolith which may therefore remain gas-tight and thus a more suitable design. These predictions are consistent with experimental findings.

Journal article

Than YR, Wenman MR, Grimes RW, 2020, Cu and Sb in tetragonal ZrO2 on fuel cladding, Journal of Applied Physics, Vol: 128, Pages: 135101-135101, ISSN: 0021-8979

Atomic scale simulations were used to predict defect formation in tetragonal ZrO2 doped with Cu and Sb. Both dopants form strong associations with oxygen vacancies impeding oxygen progression through the oxide. Sb suppresses the free oxygen vacancy population though Cu increases the concentration. Thus, while the addition of Sb is predicted to be beneficial against corrosion, Cu will show a more complex behavior. Previous simulations showed that Ni0 promotes molecular hydrogen dissociation. Neither Cu nor Sb exhibit this behavior despite Cu+ having the same electronic configuration as Ni0. Both Cu and Sb show a favorable response to applied local space charges.

Journal article

Dong P, Scatigno GG, Wenman MR, 2020, Effect of Salt Composition and Microstructure on Stress Corrosion Cracking of 316L Austenitic Stainless Steel for Dry Storage Canisters, Journal of Nuclear Materials, Pages: 152572-152572, ISSN: 0022-3115

Journal article

Harrison RW, Gasparrini C, Worth RN, Buckley J, Wenman MR, Abram Tet al., 2020, On the oxidation mechanism of U3Si2 accident tolerant nuclear fuel, CORROSION SCIENCE, Vol: 174, ISSN: 0010-938X

Journal article

Williams RJ, Vecchiato F, Kelleher J, Wenman MR, Hooper PA, Davies CMet al., 2020, Effects of heat treatment on residual stresses in the laser powder bed fusion of 316L stainless steel: Finite element predictions and neutron diffraction measurements, Journal of Manufacturing Processes, Vol: 57, Pages: 641-653, ISSN: 1526-6125

Heat treatments are used in laser powder bed fusion (LPBF) to reduce residual stress and improve service life. In order to qualify components for service, the degree of stress relaxation under heat treatment must be known. In this work, the effect of heat treatment on residual stress (RS) in LPBF 316L stainless steel was studied. Finite element (FE) models were developed to predict the RS distribution in specimens in the as-built state and subjected to heat treatment. The models simulated the thermo-mechanical LPBF build process, sample removal from the build plate and creep stress relaxation effects from a 2 h heat treatment at 700 C. The predictions were validated by neutron diffraction measurements in as-built and heat treated samples, in both build orientations. Large tensile RS of around 450 MPa were predicted at the vertical sample's outer gauge surfaces, balanced by high compressive stresses of similar magnitude at the centre. The residual stresses in the horizontal sample were significantly lower, by around 40%. The influence of sample removal from the base plate on the RS distribution was found to be strongly dependent on the sample orientation and geometry. The heat treatment preserved the unique microstructure of the LPBF process and reduced the peak RS by around 10% in the vertical sample and 40% in the horizontal sample. The FE model predictions were found in good agreement with the experimental measurements, thus providing an effective tool for RS predictions in LPBF components and proving the effectiveness of the heat treatment on RS relaxation.

Journal article

Pavlov T, lestak, Wenman M, Vlahovic, Robba, Cambriani, Staicu, Dahms, Ernstberger, Brown, Bradford M, Konings R, Grimes Ret al., 2020, Examining the thermal properties of unirradiated nuclear grade graphite between 750 and 2500 K, Journal of Nuclear Materials, Vol: 538, Pages: 1-11, ISSN: 0022-3115

This study presents the first high temperature measurements (between 750 K and 2500 K) of thermal conductivity, thermal diffusivity, specific heat and spectral emissivity of virgin graphite samples (type IM1-24) from advanced gas-cooled reactor (AGR) fuel assembly bricks. Scanning electron microscope (SEM) and X-ray computed tomography (XRT) techniques were used to verify the presence of Gilsocarbon filler particles (a characteristic microstructural feature of IM1-24 graphite). All thermal properties were investigated in two orthogonal directions, which showed the effective macroscopic thermal conductivity to be the same (to within experimental error). This can be linked to the morphology of the filler particles that consist of concentrically aligned graphitic platelets. The resulting spherical symmetry allows for heat to flow in the same manner in both macroscopic directions. The current thermal conductivity results were compared to other isotropic grade graphite materials. The significant discrepancies between the thermal conductivities of the individual grades are likely the result of different manufacturing processes yielding variations in the microstructure of the final product. Differences were identified in the filler particle size and structure, and possibly the degree of graphitization compared to other reported nuclear graphites.

Journal article

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