Materials Science › Materials Chemistry

Materials Challenges in Fusion Energy Research

Works Count: 41025

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This cluster of papers focuses on the materials challenges in fusion energy research, including radiation damage, plasma-facing components, irradiation-resistant steels, tungsten alloys, neutron irradiation effects, oxide dispersion-strengthened steels, and the development of interatomic potentials and molecular dynamics simulations for studying material behavior under fusion-relevant conditions.

Keywords

fusion materials; radiation damage; plasma-facing components; irradiation-resistant steels; tungsten alloys; neutron irradiation effects; oxide dispersion-strengthened steels; structural materials; interatomic potential; molecular dynamics simulations

Physics of Plasma-Wall Interactions in Controlled Fusion

1986-01-01

D.E. Post , R. Behrisch

Oxide dispersion-strengthened steels: A comparison of some commercial and experimental alloys

2005-03-24

R.L. Klueh , John Shingledecker , R.W. Swindeman +1 more

Recent progress in understanding reactor pressure vessel steel embrittlement

1998-06-01

G.R. Odette , G.E. Lucas

Abstract This paper reviews the current understanding of the basic mechanisms of irradiation embrittlement in reactor pressure vessel steels. Radiation enhanced diffusiona at operating temperatures around 290°C leads to the … Abstract This paper reviews the current understanding of the basic mechanisms of irradiation embrittlement in reactor pressure vessel steels. Radiation enhanced diffusiona at operating temperatures around 290°C leads to the formation of various ultrafine scale hardening phases, including copper rich and copper catalysed manganese-nickel rich precipitates. Other nanofeatures that do not require copper, so-called matrix defects, include alloy phosphides and carbonitrides as well as defect cluster-solute complexes. Matrix defects that are thermally unstable (anneal) under irradiation play a very important role in mediating flux and temperature effects. The balance of features depends on the composition of the steel and the irradiation conditions. Copper enriched phases, which are the dominant embrittling feature in alloys containing significant trace quantities of this element, are fairly well understood. In contrast, the detailed identity and etiology of the matrix defects and manganese-nickel rich phases that may form in very low copper steels has not yet been established. Embrittlement of typical (Mn-Mo-Ni) pressure vessel steels, manifested as shifts in Charpy V-notch transition temperature, can generally be related to yield stress increases. Yield stress increases from copper rich precipitates are consistent with predictions using defect-obstacle interaction theory coupled with a new model for superposition of the hardening from both pre- and post-irradiation sources of strength. Details of the strengthening contributions from the other irradiation features are not as well established, but appear to be reasonably consistent with theory. These concepts have led to the development of thermodynamic-kinetic-micromechanical models that are broadly consistent with experiment, and rationalize the highly synergistic effects of important irradiation (e.g., temperature, flux, fluence) and metallurgical (e.g., copper, nickel, manganese, phosphorous and heat treatment) variables on both irradiation hardening and hardening recovery during post-irradiation annealing. Open questions can be addressed with a hierarchy of new theoretical and experimental tools, which range from atomistic modeling to tomographic methods of observing the sequence-of-events leading to fracture. Advanced microstructural evolution, microstructure-property and micromechanical models, validated and calibrated by well designed experiments, will greatly enhance our ability to predict pressure vessel embrittlement and to resolve out-standing technical issues.

Interstitials and Vacancies in<mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline"><mml:mi>α</mml:mi></mml:math>Iron

1964-06-01

R. A. Johnson

The migration energies and atomic configurations for mono- and di-interstitials and mono- and divacancies in $\ensuremath{\alpha}$ iron have been calculated using a classical model. About 530 atoms surrounding the defect … The migration energies and atomic configurations for mono- and di-interstitials and mono- and divacancies in $\ensuremath{\alpha}$ iron have been calculated using a classical model. About 530 atoms surrounding the defect were treated as individual particles, each with three degrees of freedom, while the remainder of the crystal was treated as an elastic continuum with atoms imbedded in it. A two-body central force was devised which matched the elastic moduli, was sharply repulsive at close separation, and which went to zero midway between the second and third neighboring atoms. Configurations were found by choosing a starting configuration roughly approximating the situation under consideration and successively adjusting the value of each variable occurring in the energy equation so that the magnitude of the generalized force associated with it was zero until equilibrium was reached. The energy calculations include changes in bond energy in the discrete region, energy in the elastic field, and work done against cohesive forces, but neglect changes due to the redistribution of electrons. Calculated activation energies for motion of mono- and di-interstitials and mono- and di-vacancies were 0.33, 0.18, 0.68, and 0.66 eV, respectively, and binding energies of di-interstitials and di-vacancies were 1.08 and 0.20 eV, respectively. The stable interstitial was a "split" configuration in which two atoms were symmetrically split in a $〈110〉$ direction about a vacant normal lattice site, and the stable di-interstitial consisted of two parallel split interstitials at nearest-neighbor lattice sites with their axes perpendicular to the line joining their centers. In the vacancy configuration an atom was missing from a normal lattice site, and the divacancy consisted of two vacancies at second-nearest-neighbor lattice sites.

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Solution hardening by tetragonal dist ortions: Application to irradiation hardening in F.C.C. crystals

1962-09-01

R. L. Fleischer

Dislocation Velocities, Dislocation Densities, and Plastic Flow in Lithium Fluoride Crystals

1959-02-01

W. G. Johnston , John J. Gilman

Velocities of individual dislocations have been measured in LiF, covering a range of twelve orders of magnitude in velocity, from 10−7 cm/sec to 105 cm/sec. The velocity is extremely sensitive … Velocities of individual dislocations have been measured in LiF, covering a range of twelve orders of magnitude in velocity, from 10−7 cm/sec to 105 cm/sec. The velocity is extremely sensitive to applied stress at low velocities, and for each crystal there exists a minimum stress for dislocation motion, below which dislocations do not move. The edge components of dislocation loops move considerably faster than the screw components. The upper limit for dislocation velocity appears to be the velocity of sound in the crystal. The effects of temperature, impurities, and radiation damage on dislocation velocity are described. These variables affect the dynamic resistance to motion encountered by a moving glide dislocation. The growth of total dislocation density, the growth of individual glide bands, and the distribution of glide dislocations during plastic deformation are described. The yield stress of LiF is determined by the resistance to motion encountered by a glide dislocation in moving through an otherwise dislocation-free region of the crystal. The yield stress is independent of the dislocations present in an undeformed crystal, and the state of pinning and geometrical arrangement of such dislocations do not affect the yield stress. Stress-strain curves have been calculated from the data on dislocation mobility and dislocation density, and the calculated and measured curves are compared. At low strains the flow stress can be predicted from measured dislocation properties.

Ferritic/martensitic steels for next-generation reactors

2007-05-11

R.L. Klueh , Andrew Nelson

Present development status of EUROFER and ODS-EUROFER for application in blanket concepts

2005-08-03

R. Lindau , A. Möslang , M. Rieth +7 more

<i>Ab initio</i>study of foreign interstitial atom (C, N) interactions with intrinsic point defects in<mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline"><mml:mi>α</mml:mi></mml:math>-Fe

2004-04-19

Christophe Domain , Charlotte Becquart , J. Foct

The behavior of carbon and nitrogen atoms in iron based solid solution is studied by ab initio density-functional-theory calculations. The interaction of a C or a N atom in $\ensuremath{\alpha}$-Fe … The behavior of carbon and nitrogen atoms in iron based solid solution is studied by ab initio density-functional-theory calculations. The interaction of a C or a N atom in $\ensuremath{\alpha}$-Fe with a vacancy, other C or N interstitials as well as self-interstitial atoms is discussed and compared to known experimental results. The migration of these two foreign interstitial atoms is determined in pure Fe or when a vacancy is present in the supercell. According to our results, there is a strong binding energy of C or N with vacancies, whereas a repulsion is observed with self-interstitial atoms. Furthermore, a vacancy can trap up to two C, and a covalent bonding forms between the two C atoms. The situation is not as clear for N atoms, and a competition between the formation of N-V pairs and NN-V triplets is very probable.

Physics basis and design of the ITER plasma-facing components

2011-02-07

R.A. Pitts , S. Carpentier , F. Escourbiac +7 more

Formation process of tungsten nanostructure by the exposure to helium plasma under fusion relevant plasma conditions

2009-08-13

Shin Kajita , Wataru Sakaguchi , N. Ohno +2 more

Helium irradiation on tungsten changes the surface morphology dramatically by forming a nanometre-sized fibreform structure which could bring about serious problems for fusion reactors. From the experimental results in liner … Helium irradiation on tungsten changes the surface morphology dramatically by forming a nanometre-sized fibreform structure which could bring about serious problems for fusion reactors. From the experimental results in liner divertor simulators, it is revealed that the incident ion energy and surface temperature are key parameters for the formation of the structure. It is shown that the tungsten nanostructure is easily formed when the temperature is in the range 1000–2000 K, and the incident ion energy is higher than 20 eV. Furthermore, on the basis of the helium irradiation experiments performed in the divertor simulator NAGDIS-I, the initial formation process of the nanostructure is revealed. It is shown that the nanostructure formation is related to pinholes appearing on the bulk part of the material, and then, the rough structure develops to a much finer nanostructure. The nanostructure was also observed on the molybdenum surface that was exposed to the helium plasma. It increases interest in the possibility that nanostructure formation by helium irradiation is a common phenomenon that occurs on various metals.

Formation of Nanostructured Tungsten with Arborescent Shape due to Helium Plasma Irradiation

2006-01-01

Shuichi Takamura , N. Ohno , Dai Nishijima +1 more

Deeply nanostructured tungsten with an arborescent shape was found for the first time to be formed on tungsten-coated graphite by a high-flux helium plasma irradiation at surface temperatures of 1250 … Deeply nanostructured tungsten with an arborescent shape was found for the first time to be formed on tungsten-coated graphite by a high-flux helium plasma irradiation at surface temperatures of 1250 and 1600 K, an incident ion energy of 12 eV (well below the physical sputtering threshold) and a helium ion fluence of 3.5 × 1027 m-2.

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Perspective of ODS alloys application in nuclear environments

2002-12-01

Shigeharu Ukai , Masayuki Fujiwara

Tungsten as material for plasma-facing components in fusion devices

2011-02-12

V. Philipps

Tritium inventory in ITER plasma-facing materials and tritium removal procedures

2008-08-21

J. Roth , E. Tsitrone , T. Loarer +7 more

Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent … Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent issue in this choice is the tritium inventory build-up in the vessel, which must be limited for safety reasons. The initial material selection, i.e. beryllium (Be) on the main vessel walls, tungsten (W) on the divertor upper baffle and dome, and carbon fibre composite around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimize the lifetime of the plasma-facing components.

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Comparison of swelling and irradiation creep behavior of fcc-austenitic and bcc-ferritic/martensitic alloys at high neutron exposure

2000-01-01

F.А. Garner , M.B. Toloczko , Bulent H. Sencer

A full tungsten divertor for ITER: Physics issues and design status

2013-01-10

R.A. Pitts , S. Carpentier , F. Escourbiac +7 more

Helium induced nanoscopic morphology on tungsten under fusion relevant plasma conditions

2008-01-23

M.J. Baldwin , R. Doerner

Polished W discs exposed to pure He plasma in the PISCES-B linear-divertor-plasma simulator at 1120 and 1320 K are found to develop deeply nanostructured surface layers consisting of a conglomerate … Polished W discs exposed to pure He plasma in the PISCES-B linear-divertor-plasma simulator at 1120 and 1320 K are found to develop deeply nanostructured surface layers consisting of a conglomerate of amorphous ‘nanorods’. The growth of the thickness of the nanostructured layer is explored for exposure times spanning 300–(2.2 × 10 4 ) s in He plasmas of density n e ∼ 4 × 10 18 m −3 and temperature T e ∼ 6–8 eV where the average He-ion surface-impact energy is ∼60 eV, below the threshold for physical sputtering. A nanostructured layer in excess of 5 µm thick is observed for the longest exposure time explored. The kinetics of the layer growth are found to follow Fick's law, characterized by an effective diffusive mechanism with coefficients of diffusion: D 1120 K = 6.6 ± 0.4 × 10 −12 cm 2 s −1 and D 1320 K = 2.0± 0.5 × 10 −11 cm 2 s −1 . The diffusion of He atoms in W is considered too rapid to explain the observed growth of surface modification and points to the interplay of other mechanisms, such as the availability of thermal vacancies and/or the slower diffusion of He through the forming nanostructured layer.

On the stress dependence of partial dislocation separation and deformation microstructure in austenitic stainless steels

2003-04-24

Thak Sang Byun

Recent progress in research on tungsten materials for nuclear fusion applications in Europe

2012-08-28

M. Rieth , S. L. Dudarev , S.M. González de Vicente +7 more

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Modified embedded atom potentials for HCP metals

1994-01-01

M. I. Baskes , R. A. Johnson

The modified embedded atom method (MEAM) is an empirical extension of embedded atom method (EAM) that includes angular forces. The MEAM, which has previously been applied to the atoms in … The modified embedded atom method (MEAM) is an empirical extension of embedded atom method (EAM) that includes angular forces. The MEAM, which has previously been applied to the atoms in the FCC, BCC, and diamond cubic crystal systems, has been extended to the HCP crystal structure. Parameters have been determined for HCP metals that have c/a ratios less than ideal. The model is fitted to the lattice constants, elastic constants, cohesive energy, vacancy formation energy, and the BCC-HCP structural energy difference of these metals and is able to reproduce this extensive data base quite well. Structural energies and lattice constants of the HCP metals in a number of cubic structures are predicted. The divacancy is found to be unbound in all of the metals considered except for Be. Stacking fault and surface energies are found to be in reasonable agreement with experiment.

Efficient Annealing of Radiation Damage Near Grain Boundaries via Interstitial Emission

2010-03-25

Xian-Ming Bai , Arthur F. Voter , Richard G. Hoagland +2 more

Although grain boundaries can serve as effective sinks for radiation-induced defects such as interstitials and vacancies, the atomistic mechanisms leading to this enhanced tolerance are still not well understood. With … Although grain boundaries can serve as effective sinks for radiation-induced defects such as interstitials and vacancies, the atomistic mechanisms leading to this enhanced tolerance are still not well understood. With the use of three atomistic simulation methods, we investigated defect-grain boundary interaction mechanisms in copper from picosecond to microsecond time scales. We found that grain boundaries have a surprising "loading-unloading" effect. Upon irradiation, interstitials are loaded into the boundary, which then acts as a source, emitting interstitials to annihilate vacancies in the bulk. This unexpected recombination mechanism has a much lower energy barrier than conventional vacancy diffusion and is efficient for annihilating immobile vacancies in the nearby bulk, resulting in self-healing of the radiation-induced damage.

Multiscale modelling of defect kinetics in irradiated iron

2004-12-19

Chu‐Chun Fu , J. Dalla Torre , F. Willaime +2 more

Vacancy defect mobilities and binding energies obtained from annealing studies

1978-02-01

R. W. Balluffi

Helium accumulation in metals during irradiation – where do we stand?

2003-11-15

H. Trinkaus , B.N. Singh

Interaction of electrons with lattice vibrations

1952-12-05

H. Fröhlich

Using methods of modern field theories a canonical transformation of the Hamiltonian of free electrons in the field of the lattice vibrations is performed. This transformation takes account of the … Using methods of modern field theories a canonical transformation of the Hamiltonian of free electrons in the field of the lattice vibrations is performed. This transformation takes account of the bulk of the interaction of the electrons with the vibrational field and leads to a renormalization of the velocity of sound and of the interaction parameter F . An objection of Wentzel’s against the use of large F is removed in this way. Even in the case of weak interaction the transformed Hamiltonian contains already in zero order terms which require a modification of the usual procedure in the theory of metals, and which at low temperatures lead to an increase of the effective mass of the electrons. Treatment of strong interaction requires the development of a new method.

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Low-activation ferritic and martensitic steels for fusion application

1996-10-01

A. Kohyama , A. Hishinuma , D.S. Gelles +3 more

On the role of alloy composition and processing parameters in nanocluster formation and dispersion strengthening in nanostuctured ferritic alloys

2008-10-28

M.J. Alinger , G.R. Odette , D. Hoelzer

The relationship between hardness and yield stress in irradiated austenitic and ferritic steels

2004-12-13

Jeremy T. Busby , M.C. Hash , Gary S. Was

Recent analysis of key plasma wall interactions issues for ITER

2009-01-27

Joachim Roth , E. Tsitrone , A. Loarte +7 more

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Plasma facing and high heat flux materials – needs for ITER and beyond

2002-12-01

H. Bolt , V. Barabash , G. Federici +4 more

Microstructure evolution in tungsten during low-energy helium ion irradiation

2000-12-01

Hirotomo Iwakiri , K Yasunaga , Kazunori Morishita +1 more

<i>Ab initio</i>study of helium in<mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline"><mml:mrow><mml:mi>α</mml:mi><mml:mtext>−</mml:mtext><mml:mi>Fe</mml:mi></mml:mrow></mml:math>: Dissolution, migration, and clustering with vacancies

2005-08-24

Chu‐Chun Fu , F. Willaime

Density functional theory calculations have been performed to study the dissolution and migration of helium in $\ensuremath{\alpha}$-iron, and the stability of small helium-vacancy clusters ${\mathrm{He}}_{n}{V}_{m}$ ($n$,$m=0$ to 4). Substitutional and … Density functional theory calculations have been performed to study the dissolution and migration of helium in $\ensuremath{\alpha}$-iron, and the stability of small helium-vacancy clusters ${\mathrm{He}}_{n}{V}_{m}$ ($n$,$m=0$ to 4). Substitutional and interstitial configurations of helium are found to have similar stabilities. The tetrahedral configuration is more stable than the octahedral by 0.2 eV. Interstitial helium atoms are predicted to have attractive interactions and a very low migration energy (0.06 eV), suggesting that He bubbles can form at low temperatures in initially vacancy-free lattices. The migration of substitutional helium by the vacancy mechanism is governed by the migration of the ${\mathrm{HeV}}_{2}$ complex, with an energy barrier of 1.1 eV. The activation energies for helium diffusion by the dissociation and vacancy mechanisms are estimated for the limiting cases of thermal-vacancy regime and of high supersaturation of vacancies. The trends of the binding energies of vacancy and helium to helium-vacancy clusters are discussed in terms of providing additional knowledge on the behavior of He in irradiated iron, necessary for the interpretation of complex experimental data such as thermal He desorption spectra.

Hydrogen isotope retention and recycling in fusion reactor plasma-facing components

2002-02-01

R.A. Causey

Operating temperature windows for fusion reactor structural materials

2000-11-01

S.J. Zinkle , N.M. Ghoniem

The evolution of mechanical property change in irradiated austenitic stainless steels

1993-11-01

G.E. Lucas

Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

1993-11-01

S.J. Zinkle , P.J. Maziasz , R.E. Stoller

Designing Radiation Resistance in Materials for Fusion Energy

2014-04-04

S.J. Zinkle , L.L. Snead

Proposed fusion and advanced (Generation IV) fission energy systems require high-performance materials capable of satisfactory operation up to neutron damage levels approaching 200 atomic displacements per atom with large amounts … Proposed fusion and advanced (Generation IV) fission energy systems require high-performance materials capable of satisfactory operation up to neutron damage levels approaching 200 atomic displacements per atom with large amounts of transmutant hydrogen and helium isotopes. After a brief overview of fusion reactor concepts and radiation effects phenomena in structural and functional (nonstructural) materials, three fundamental options for designing radiation resistance are outlined: Utilize matrix phases with inherent radiation tolerance, select materials in which vacancies are immobile at the design operating temperatures, or engineer materials with high sink densities for point defect recombination. Environmental and safety considerations impose several additional restrictions on potential materials systems, but reduced-activation ferritic/martensitic steels (including thermomechanically treated and oxide dispersion–strengthened options) and silicon carbide ceramic composites emerge as robust structural materials options. Materials modeling (including computational thermodynamics) and advanced manufacturing methods are poised to exert a major impact in the next ten years.

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

2001-12-01

G. Federici , C.H. Skinner , J.N. Brooks +7 more

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety … The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Recent Developments in Irradiation-Resistant Steels

2008-05-08

G.R. Odette , M.J. Alinger , Brian D. Wirth

Advanced fission and future fusion energy will require new high-performance structural alloys with outstanding properties that are sustained under long-term service in ultrasevere environments, including neutron damage producing up to … Advanced fission and future fusion energy will require new high-performance structural alloys with outstanding properties that are sustained under long-term service in ultrasevere environments, including neutron damage producing up to 200 atomic displacements per atom and, for fusion, 2000 appm of He. Following a brief description of irradiation damage and damage resistance, we focus on an emerging class of nanostructured ferritic alloys (NFAs) that show promise for meeting these challenges. NFAs contain an ultrahigh density of Y-Ti-O-enriched dispersion-strengthening nanofeatures (NFs) that, along with fine grains and high dislocation densities, provide remarkably high tensile, creep, and fatigue strength. The NFs are stable under irradiation up to 800°C and trap He in fine-scale bubbles, suppressing void swelling and fast fracture embrittlement at lower temperatures and creep rupture embrittlement at high temperatures. The current state of the development and understanding of NFAs is described, along with some significant outstanding challenges.

Elevated temperature ferritic and martensitic steels and their application to future nuclear reactors

2005-09-27

R.L. Klueh

In the 1970s, high chromium (9–12%Cr) ferritic/martensitic steels became candidates for elevated temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, … In the 1970s, high chromium (9–12%Cr) ferritic/martensitic steels became candidates for elevated temperature applications in the core of fast reactors. Steels developed for conventional power plants, such as Sandvik HT9, a nominally Fe–12Cr–1Mo–0·5W–0·5Ni–0·25V–0·2C steel (composition in wt-%), were considered in the USA, Europe and Japan. Now, a new generation of fission reactors is in the planning stage, and ferritic, bainitic and martensitic steels are again candidates for in-core and out-of-core applications. Since the 1970s, advances have been made in developing steels with 2–12%Cr for conventional power plants that are significant improvements over steels originally considered. The present study will review the development of the new steels to illustrate the advantages they offer for the new reactor concepts. Elevated temperature mechanical properties will be emphasised. Effects of alloying additions on long-time thermal exposure with and without stress (creep) will be examined. Information on neutron radiation effects will be discussed as it applies to ferritic and martensitic steels.

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Design of Radiation Tolerant Materials Via Interface Engineering

2013-09-24

Wei‐Zhong Han , Michael J. Demkowicz , Nathan A. Mara +7 more

A novel interface engineering strategy is proposed to simultaneously achieve superior irradiation tolerance, high strength, and high thermal stability in bulk nanolayered composites of a model face-centered-cubic (Cu)/body-centered-cubic (Nb) system. … A novel interface engineering strategy is proposed to simultaneously achieve superior irradiation tolerance, high strength, and high thermal stability in bulk nanolayered composites of a model face-centered-cubic (Cu)/body-centered-cubic (Nb) system. By synthesizing bulk nanolayered Cu-Nb composites containing interfaces with controlled sink efficiencies, a novel material is designed in which nearly all irradiation-induced defects are annihilated. As a service to our authors and readers, this journal provides supporting information supplied by the authors. Such materials are peer reviewed and may be re-organized for online delivery, but are not copy-edited or typeset. Technical support issues arising from supporting information (other than missing files) should be addressed to the authors. Please note: The publisher is not responsible for the content or functionality of any supporting information supplied by the authors. Any queries (other than missing content) should be directed to the corresponding author for the article.

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Analytic embedded atom method model for bcc metals

1989-10-01

R. A. Johnson , Donghwan Oh

Computer simulation of point defect properties in dilute Fe—Cu alloy using a many-body interatomic potential

1997-03-01

Graeme J. Ackland , David Bacon , A.F. Calder +1 more

Abstract The behaviour of copper atoms in dilute solution in α-iron is important for the microstructural changes that occur in ferritic pressure vessel steels under fastneutron irradiation. To investigate the … Abstract The behaviour of copper atoms in dilute solution in α-iron is important for the microstructural changes that occur in ferritic pressure vessel steels under fastneutron irradiation. To investigate the properties of atomic defects that control this behaviour, a set of many-body interatomic potentials has been developed for the Fe—Cu system. The procedures employed, including modifications to ensure suitability for simulating atomic collisions at high energy, are described. The effect of copper on the lattice parameter of iron in the new model is in good agreement with experiment. The phonon properties of the pure crystals and, in particular, the influence of the instability of the metastable, bcc phase of copper that precipitates during irradiation are discussed. The properties of point defects have been investigated. It is found that the vacancy has lower formation and migration energy in bcc copper than in α-iron, and the self-interstitial atom has very low formation energy in this phase of copper. The threshold displacement energy in iron has been computed as a function of recoil orientation for both iron-and copper-atom recoils. The differences between the energy for the two species are small.

The primary damage state in fcc, bcc and hcp metals as seen in molecular dynamics simulations

2000-01-01

David Bacon , Fei Gao , Yu.N. Osetsky

Materials for the plasma-facing components of fusion reactors

2004-06-21

H. Bolt , V. Barabash , W. Krauss +5 more

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Mechanism of Radiation Damage Reduction in Equiatomic Multicomponent Single Phase Alloys

2016-04-01

F. Granberg , K. Nordlund , Mohammad W. Ullah +7 more

Recently a new class of metal alloys, of single-phase multicomponent composition at roughly equal atomic concentrations (``equiatomic''), have been shown to exhibit promising mechanical, magnetic, and corrosion resistance properties, in … Recently a new class of metal alloys, of single-phase multicomponent composition at roughly equal atomic concentrations (``equiatomic''), have been shown to exhibit promising mechanical, magnetic, and corrosion resistance properties, in particular, at high temperatures. These features make them potential candidates for components of next-generation nuclear reactors and other high-radiation environments that will involve high temperatures combined with corrosive environments and extreme radiation exposure. In spite of a wide range of recent studies of many important properties of these alloys, their radiation tolerance at high doses remains unexplored. In this work, a combination of experimental and modeling efforts reveals a substantial reduction of damage accumulation under prolonged irradiation in single-phase NiFe and NiCoCr alloys compared to elemental Ni. This effect is explained by reduced dislocation mobility, which leads to slower growth of large dislocation structures. Moreover, there is no observable phase separation, ordering, or amorphization, pointing to a high phase stability of this class of alloys.

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Materials research for fusion

2016-05-01

J. Knaster , A. Moeslang , T. Muroga

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Physics basis for the first ITER tungsten divertor

2019-07-18

R.A. Pitts , X. Bonnin , F. Escourbiac +7 more

• Reviews the fundamental physics aspects of the first ITER W divertor and defines the required operational lifetime within the Staged Approach. • Uses the ITER divertor SOLPS simulation database … • Reviews the fundamental physics aspects of the first ITER W divertor and defines the required operational lifetime within the Staged Approach. • Uses the ITER divertor SOLPS simulation database to establish the target peak heat flux and neutral pressure burning plasma operating domain. • Assesses consequences of narrow SOL heat flux channels, fluid drifts, component shaping and 3D magnetic fields for ELM control. • Uses W recrystallization to define an operational budget and shows that heat fluxes ∼50% higher than previously assumed may be acceptable. • Shows that Ne and N should be equally good as seed impurities and suggests that very strong ELM mitigation will be required at high performance. • Provides a list of key outstanding R&D areas to consolidate the divertor physics basis in the period up to ITER operation. On the eve of component procurement, this paper discusses the present physics basis for the first ITER tungsten (W) divertor, beginning with a reminder of the key elements defining the overall design, and outlining relevant aspects of the Research Plan accompanying the new "staged approach" to ITER nuclear operations which fixes the overall divertor lifetime constraint. The principal focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SOLPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne). A new perspective on the simulation database is adopted, concentrating purely on the divertor physics aspects rather than on the core-edge integration, which has been studied extensively in the course of the divertor design evolution and is published elsewhere. Emphasis is placed on factors which may increase the peak steady state loads: divertor target shaping for component misalignment protection, the influence of fluid drifts, and the consequences of narrow scrape-off layer heat flux channels. All tend to push the divertor into an operating space at higher sub-divertor neutral pressure in order to remain at power flux densities acceptable for the target material. However, a revised criterion for the maximum tolerable loads based on avoidance of W recrystallization, sets an upper limit potentially ∼50% higher than the previously accepted value of ∼10 MW m −2 , a consequence both of the choice of material and the finalized component design. Although the simulation database is currently restricted to the 2-D toroidally symmetric situation, considerable progress is now also being made using the EMC3-Eirene 3-D code suite for the assessment of power loading in the presence of magnetic perturbations for ELM control. Some new results for low input power corresponding to the early H-mode operation phases are reported, showing that even if realistic plasma screening is taken into account, significant asymmetric divertor heat fluxes may arise far from the unperturbed strike point. The issue of tolerable limits for transient heat pulses is an open and key question. A new scaling for ELM power deposition has shown that whilst there may be more latitude for operation at higher current without ELM control, the ultimate limit is likely to be set more by material fatigue under large numbers of sub-threshold melting events.

High-Chromium Ferritic and Martensitic Steels for Nuclear Applications

2001-01-01

R.L. Klueh , Dennis Harries

Description The desire to reduce generating costs by increasing thermal efficiency in power plants has resulted in a need for materials with improved high-temperature properties. Advanced ferritic/martensitic steels are prime … Description The desire to reduce generating costs by increasing thermal efficiency in power plants has resulted in a need for materials with improved high-temperature properties. Advanced ferritic/martensitic steels are prime candidates for such applications. Although these alloys were developed for application in nuclear power plants because of their low swelling behavior, their high creep-rupture strength and appealing thermal properties have led to their use in other demanding environments. Monograph 3 provides a detailed review of the development of the high-chromium ferritic/martensitic steels for exposure to the high-energy neutron environment of a fission or fusion reactor. Since the steels are used extensively for a range of high-temperature applications, this book also provides insight into the basic properties of the steels under non-nuclear conditions. The many materials scientists worldwide who conduct research on non-nuclear applications will find this a valuable resource. A comprehensive list of references is also provided and will save users valuable time otherwise spent on locating related publications.

Development and demonstration of basics for plasma reprocessing of spent nuclear fuel

2025-06-25

Н. Н. Антонов , M.A. Valinurov , А. В. Гавриков +7 more | Atomic Energy

“Canode”: A conical partially magnetic anode for efficient negative ion extraction from duoplasmatron ion sources

2025-06-24

Peter Williams , Maitrayee Bose , R. L. Hervig +2 more | Journal of Vacuum Science & Technology B Nanotechnology and Microelectronics Materials Processing Measurement and Phenomena

We report on the design and performance of an improved duoplasmatron ion source for secondary ion mass spectrometers. The source is designed specifically to optimize extraction of negative oxygen ions … We report on the design and performance of an improved duoplasmatron ion source for secondary ion mass spectrometers. The source is designed specifically to optimize extraction of negative oxygen ions while suppressing electron extraction using a built-in magnetic asymmetry in the anode electrode. Other changes from conventional designs are (a) drilling the ion extraction aperture directly into the magnetic steel anode rather than in a refractory (nonmagnetic) metal insert, thereby eliminating a magnetic “hole” that acts to counter the desired magnetic concentration of the discharge at the aperture and (b) forming the anode into a conical shape convex toward the intermediate electrode to increase the magnetic field concentration at the extraction aperture, hence the term “Canode.” The built-in magnetic asymmetry allows the width and shape of the intermediate electrode to be varied to further optimize magnetic concentration of the discharge. Tests were performed with both ims 6f and NanoSIMS 50L instruments manufactured by Cameca Instruments, Inc. (Fitchburg, WI, USA). In the ims 6f, the Canode design gave O− primary ion currents up to a factor of five greater than the factory ion source design. In the NanoSIMS 50L, the Canode source produced a focused O− ion beam at the sample with a diameter of 50 nm, identical to the performance of the radio-frequency Hyperion ion source developed by Oregon Physics (Beaverton, OR, USA) and offered as an option by Cameca.

Nanopore-assisted rapid Cr diffusion in irradiated additive-manufactured steel

2025-06-23

Jiaying Zhou , Siyi Qiu , Mengmeng Yang +6 more | Journal of Iron and Steel Research International

FC-FNS Code for Simulating Distributions and Inventories of Hydrogen Isotopes in Fuel Cycle Systems of Fusion Reactor

2025-06-23

S.S. Ananyev , B. V. Kuteev | Fusion Science & Technology

Defect formation and irradiation response in Ti-5331 alloy under helium ion exposure

2025-06-20

Zhenping Wu , Yunmei Shi , Qigui Yang +7 more | Vacuum

Advancements in the IFMIF-DONES target system for fusion materials irradiation

2025-06-20

D. Oravecz , A. Zsákai , R.L. Csiszár +6 more | Fusion Engineering and Design

Optimizing Hydrogen Resistance in Fusion Energy Structural Stainless Steel via Controlled Uniaxial Strain

2025-06-19

L.Y. Mao , Manquan Fang , Chongyu Shen +4 more | Materialia

A study of helium bubble facetation in bcc iron through molecular dynamics

2025-06-19

Zhengyang Ming , Ze Chen , Zhaofan Wang +5 more | Scripta Materialia

Phase Evolution and Synthesis of Be12 Nb Intermetallic Compound in the 800–1300 °C Temperature Range

2025-06-19

Sergey Udartsev , Inesh Kenzhina , Timur Kulsartov +7 more | Materials

Beryllium-based intermetallic compounds, such as Be12Nb, are attracting growing interest for their high thermal stability and potential to replace pure beryllium as neutron reflectors and multipliers in both fission and … Beryllium-based intermetallic compounds, such as Be12Nb, are attracting growing interest for their high thermal stability and potential to replace pure beryllium as neutron reflectors and multipliers in both fission and future fusion reactors, with additional applications in metallurgy, aerospace, and hydrogen technology. The paper presents the results of an investigation of the thermal treatment and phase formation of the intermetallic compound Be12Nb from a mixture of niobium and beryllium powders in the temperature range of 800-1300 °C. The phase evolution was assessed as a function of sintering temperature and time. A nearly single-phase Be12Nb composition was achieved at 1100 °C, while decomposition into lower-order beryllides such as Be17Nb2 occurred at temperatures ≥1200 °C, indicating thermal instability of Be12Nb under vacuum. Careful handling of sintering in low vacuum minimized oxidation, though signs of possible BeO formation were noted. The findings complement and extend earlier reports on Be12Nb synthesis via plasma sintering, mechanical alloying, and other powder metallurgy routes, providing broader insight into phase formation and synthesis. These results provide a foundation for optimizing the manufacturing parameters required to produce homogeneous Be12Nb-based components and billets at an industrial scale. Additionally, they help define the operational temperature limits necessary to preserve the material's phase integrity during application.

The roles of fuzzy W surfaces in avoiding W cracking during ELMs-like heat loads

2025-06-18

Weifeng Liu , Xu Zhao , Chunjie Niu +2 more | Fusion Engineering and Design

Tritium transfer from neutron-irradiated lithium to yttrium and gas phase

2025-06-18

Yutaro Mori , Kazunari Katayama , Junnosuke Haruta +7 more | Fusion Engineering and Design

Design and development of divertor pipe welding tool for fusion reactor

2025-06-18

Lei Zheng , Wanjin Zhang , Yongjun Sun +6 more | Fusion Engineering and Design

Numerical simulation of flow and heat transfer characteristics of liquid lithium metal inside trapezoid shaped wire rod bundle channels

2025-06-18

Wenchao Zhang , Qiaochen Du , Yue Wang +4 more | Annals of Nuclear Energy

Irradiation effects on additively manufactured porous 316H stainless steel: A molecular dynamics study

2025-06-17

Mahmoud A. Mahrous , Muhammad A. Abdelghany , H. Farag +1 more | Computational Materials Science

Cataloging Legacy Data from the Tritium Systems Test Assembly Program

2025-06-16

Vicky Lizardi-Lobb , Claire L. Bullock , Victoria Hypes-Mayfield +1 more | Fusion Science & Technology

Hydraulic analysis of the DTT divertor module

2025-06-14

Davide Caprini , Morena Angelucci , Emanuela Martelli +6 more | Fusion Engineering and Design

Validation of DYON simulations and development of physical sputtering models for lithiation and boronisation in EAST

2025-06-13

Runze Chen , Hyun-Tae Kim , Wenbin Liu +7 more | Nuclear Fusion

Abstract This paper validates for the first time the predictive capability of the DYON code for plasma initiation in EAST, which has metallic wall, superconducting coils and conventional tokamak shape, … Abstract This paper validates for the first time the predictive capability of the DYON code for plasma initiation in EAST, which has metallic wall, superconducting coils and conventional tokamak shape, like ITER. The model accurately reproduced the operating spaces of loop voltage and prefill gas pressure for ohmic discharges, demonstrating its validity in predicting the required operating parameters for successful inductive plasma initiation in EAST.&#xD; The role of wall conditioning on plasma initiation was investigated with the newly developed physical sputtering models of Boron and Lithium. &#xD; In EAST experiments, it was observed that the discharges after boronisation of the wall are much more vulnerable to plasma burn-through failure than after lithium-coating. The simulation results revealed that despite the similar physical sputtering yield in Boron and Lithium, the radiative energy loss rates for the boron-coated wall is significantly higher than that for the lithium-coated wall, due to the much higher radiative power coefficients of Boron.&#xD; Parametric scans of initial Boron content in ohmic discharge at the typical prefill gas pressure in EAST (0.8mPa) showed that even 1.5% of Boron content in the prefilled gas, possibly remaining after boronisation of the wall, could lead to excessive radiation energy losses and failure of plasma burn-through. For successful plasma burn-through with 1.5% initial boron content, the modelling indicates 10kW absorption of EC power is required, and it increases with more initial boron e.g. 50kW for 3% initial boron content.

Integrated modeling of RF-Induced Tungsten erosion at ICRH Antenna structures in the WEST tokamak

2025-06-13

Atul Kumar , W. Tierens , Tim Younkin +7 more | Nuclear Fusion

Abstract This paper introduces STRIPE (Simulated Transport of RF Impurity Production and Emission), an advanced modeling framework designed to analyze material erosion and the global transport of eroded impurities originating … Abstract This paper introduces STRIPE (Simulated Transport of RF Impurity Production and Emission), an advanced modeling framework designed to analyze material erosion and the global transport of eroded impurities originating from radio-frequency (RF) antenna structures in magnetic confinement fusion devices. STRIPE integrates multiple computational tools, each addressing different levels of physics fidelity: SolEdge3x for scrape-off-layer plasma profiles, COMSOL for 3D RF rectified voltage fields, RustBCA code for erosion yields and surface interactions, and GITR for 3D ion energy-angle distributions and global impurity transport. The framework is applied to an ion cyclotron RF-heated, L-mode discharge #57877 in the WEST Tokamak, where it predicts a tenfold increase in tungsten erosion at RF antenna limiters under RF-sheath rectification conditions, compared to cases with only a thermal sheath. Oxygen ions with charge states of O6+ and higher are the dominant contributors to tungsten sputtering at the antenna limiters. To verify the model's accuracy, a synthetic diagnostic tool, based on inverse photon efficiency (S/XB coefficients) from the ColRadPy collisional radiative model, enables direct comparisons between simulation results and experimental spectroscopic data. Model predictions, based on a plasma composition of 1% oxygen and 99% deuterium, show reasonable agreement with measured neutral tungsten (W-I) spectroscopic data for discharge #57877, thereby validating the accuracy of the framework. Currently, the STRIPE framework is being extended to investigate plasma-material interactions in other RF-heated linear and toroidal devices, offering valuable insights for RF antenna design, impurity control, and performance optimization in future fusion reactors.

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Mitigated Dissimilar Mass Transfer with Aluminized RAFM Steels in Flowing PbLi at 650°C

2025-06-12

Claude de Lamater-Brotherton , Marie Romedenne , Ying Zhang +1 more | Fusion Science & Technology

Properties of non-cryogenic DTs and their relevance for fusion

2025-06-03

H. Ruhl , Christian Bild , Ondrej Jaura +4 more

In inertial confinement fusion, pure deuterium-tritium (DT) is usually used as a fusion fuel. In their paper, S. Y. Guskov et al. [Plasma Phys. Rep. 37, 1020 (2011)] instead propose … In inertial confinement fusion, pure deuterium-tritium (DT) is usually used as a fusion fuel. In their paper, S. Y. Guskov et al. [Plasma Phys. Rep. 37, 1020 (2011)] instead propose using low-Z compounds that contain DT and are non-cryogenic at room temperature. They suggest that these fuels can be ignited for ρDTR≥0.35gcm−2 and kTe≥14keV, i.e., parameters that are more stringent but still in the same order of magnitude as those for DT. In deriving these results, Guskov et al. assume that ionic and electronic temperatures are equal and consider only electronic stopping power. Here, we show that at temperatures greater than 10 keV, ionic stopping power is not negligible compared to the electronic one. We demonstrate that this necessarily leads to higher ionic than electronic temperatures. Both factors facilitate ignition, showing that non-cryogenic DT compounds are more versatile than previously known. In addition, we find that heavy beryllium borohydride ignites more easily than heavy beryllium hydride, the best-performing fuel found by Guskov et al. Our results are based on an analytical model that incorporates a detailed stopping power analysis, as well as on numerical simulations using an improved version of the community hydro code MULTI-IFE. Alleviating the constraints and costs of cryogenic technology and the fact that non-cryogenic DT fuels are solids at room temperature opens up new design options for fusion targets with Q&gt;100. The discussion presented here generalizes the analysis of fuels for energy production.

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Exploratory thermal and fluid-dynamic assessment of a pin of the Helium cooled pebble bed breeding blanket concept with a modular pin monoblock design for the EU DEMO nuclear fusion reactor

2025-06-02

Alberto Gioè , Francisco A. Hernández , G. Bongiovì +7 more | Fusion Engineering and Design

Impact of heat load on tungsten material by hydrogen plasma stream of PPA

2025-06-02

A. Ahmed , S. Singha , Pranjal K. Baruah +3 more | Vacuum

THERMODYNAMICS OF HYDROGEN ISOTOPES IN TUNGSTEN

2025-06-01

N. N. Degtyarenko , К. С. Гришаков | Physics Letters A

A Data-Driven Machine Learning Model for Radiation-Induced DBTT Shifts in RAFM Steels

2025-06-01

Pengxin Wang , G.M.A.M. El-Fallah | Journal of Nuclear Materials

Corrigendum to ‘Charpy V-notch energy prediction for irradiated reactor pressure vessel steel using the extreme gradient boost method’ [Int. J. Pres. Ves. Pip. Volume 218, Part A, December 2025, 105559]

2025-06-01

Yoshinori Hashimoto , Mark Kirk | International Journal of Pressure Vessels and Piping

Simulation of tritium release processes from lithium ceramics under reactor irradiation conditions with varying reactor power

2025-06-01

Timur Kulsartov , Inesh Kenzhina , Yergazy Kenzhin +7 more | Journal of Nuclear Materials

Suppression of irradiation defects and crack propagation in niobium via grain boundary engineering: A deep potential molecular dynamics study

2025-06-01

Jiahang Li , Yajun Zhang , Jie Wang +1 more | Materials & Design

A laser-based annealing methodology to speed-up the study of thermo-activated restoration mechanisms in metals

2025-06-01

Maxime Lemetais , Jean-Baptiste Pruvo , Marco Minissale +7 more | Review of Scientific Instruments

The properties of polycrystalline materials are significantly influenced by annealing treatments. This article introduces a laser-annealing method that facilitates the investigation of high-temperature transformations, with a specific focus on tungsten … The properties of polycrystalline materials are significantly influenced by annealing treatments. This article introduces a laser-annealing method that facilitates the investigation of high-temperature transformations, with a specific focus on tungsten restoration. The aim of this research is to establish a controlled temperature gradient in the examined sample to expedite the study of restoration at high temperatures by reducing the number of samples. To achieve this, simulations are employed to design the desired temperature profile, and a laser-based setup is adapted to generate and regulate this profile. Furthermore, uncertainties and errors associated with temperature measurements in the experimental setup are quantified. The proposed laser-annealing method enables precise temperature control during the annealing process. By heating one side of a tungsten rod using the laser system, a steady-state temperature gradient is induced. The annealing process consists of two steps: the initial heating phase to reach the desired temperature profile, followed by the maintenance of constant temperatures at specific positions along the rod for a defined duration. The study investigates the impact of absorbed power by the sample on the temperature profile and assesses the softening of tungsten after annealing using hardness measurements. Overall, the proposed laser-annealing method offers a promising avenue for advancing material science research. Its ability to precisely control temperature gradients and observe material behaviors at high temperatures opens up new opportunities to optimize the properties of polycrystalline materials beyond tungsten, thus providing broader applications in material engineering and manufacturing.