Displacement Per Atom Research Articles

Attenuation of atomic displacement damage in the heavy reflector of the PERLE experiment and application to EPR

The investigation of atomic displacement damage is important for studying the irradiated materials. The present work proposes two equations to describe the total neutron-induced and gamma-induced Displacement per Atom (DPA) rates versus the penetration in materials. The attenuation of DPA rate in materials is studied in the heavy reflector of the PERLE experiment. The proposed analytic formula for neutron-induced atomic displacement well describes the Monte Carlo simulated results with two fitting parameters, while the formula for gamma-induced DPA is in agreement with Monte Carlo simulations with potential 10% uncertainties. It is noticeable that the neutron flux above 0.5 MeV is more representative than that above 0.1 MeV or 1 MeV for the attenuation of DPA in the heavy reflector. Due to the representativity of the PERLE experiment for the heavy reflector of Evolutionary Pressurized Reactor (EPR), an analytic expression of displacement damage is obtained. The atomic displacement rate at the inner surface of EPR heavy reflector is 1.4 DPA/year (using the standard Norgett-Robinson-Torrens (NRT)-DPA metric). The DPA rate at the outer surface of the reflector is reduced by a factor of 9 (and 32) as the thickness of the steel increases from 2 cm to 20 cm (and 30 cm).

Calculation and verification of neutron irradiation damage with differential cross sections

The Displacement per Atom (DPA) rate is conventionally computed with DPA cross sections in reactor applications. The method of direct calculation with energy-angular distributions given in the Center of Mass (CM) frame is proposed and recommended in the present work. The methods for refining and verifying the calculations of DPA cross sections are proposed: (i) Gauss-Legendre-Quadrature-based Piecewise Integration (GLQPI) for ensuring the numeric convergence of integral over emission angle due to the discontinuity of integrand; (ii) verification of the convergence for trapezoidal integration over the secondary energy; (iii) interpolation of double-differential cross sections. For 56Fe of JEFF-3.1.1, the current numeric integration over emission angle is shown not convergent, whereas the direct trapezoidal over the secondary energy and the direct interpolation of energy-angle-integrated damage are shown accurate. On the other hand, it is shown that the DPA cross sections are overestimated if isotropic angular distributions are assumed. However, the DPA cross section is not sensitive to the high-order Legendre polynomials because the former is an angle-integrated quantity. Numerical results of neutron elastic scattering show that 2 orders of Legendre polynomials can give the DPA rates of 56Fe within 0.5% overestimation for fission reactors, while 4 orders are required for fusion reactors. For neutron inelastic scatterings-induced DPA, the first order Legendre polynomial is sufficient for both fission and fusion reactors.

Cascade size and dose rate effects on α′ precipitation in ion-irradiated Fe14Cr alloy

Dose rate and cascade size effects on α′ formation in Fe14Cr have been studied under ion irradiation using analytical scanning transmission electron microscopy. Samples were irradiated with 100 keV He, 350 keV Ne or 1 MeV Kr ions at 300 °C to a dose of 2.5 DPA at dose rates of 10−4 and 10−3 displacements per atom (DPA)/s. Higher dose rates commensurately accelerate α′ precipitation under He ion irradiation, resulting in similar precipitate sizes and density. Number densities of α′ precipitates decreased with increasing ion mass and energy attributed to ballistic mixing from high-energy-density cascade formation and sub-cascade overlap.

Calculation and analysis of gamma-induced irradiation damage cross section

Due to the discrepancies of gamma-induced Displacement per Atom (DPA) cross sections among different calculations, the present work identifies the sources of the discrepancies and recommends the computation of displacement cross sections induced by photons, including the stopping power for electrons and positrons, the differential scattering cross sections, and the gamma-matter interactions. The production of electrons and positrons and the corresponding gamma-induced damage cross sections computed with the above methods are validated against Monte Carlo simulations. The damage cross section for Photoelectric Effect (PE) is negligible for incident photon energy higher than 2 MeV. Above 2 MeV, the percentage of atomic displacement for Compton Scattering (CS) decreases with incident energy, while that for Pair Production (PP) increases and becomes more important than CS above 10 MeV. In the PP, it is recommended to differently treat positrons and electrons for damage calculations because positrons produce 68–77% of DPA induced by electrons. 9.8 × 10−10 DPA/year gamma-induced displacement is obtained in the heavy reflector of PERLE experiment using the recommended damage cross sections. Neglecting the influence of moderator on photon flux in reflector, 8.1 × 10−4 DPA/year gamma-induced displacement damage is found in the reflector of Evolutionary Pressurized Reactor (EPR).

Study on the self-shielding and temperature influences on the neutron irradiation damage calculations in reactors

Displacement per Atom (DPA) is conventionally computed using the DPA cross sections in reactor applications. The DPA cross sections can be influenced by temperature through the Primary Knocked-on Atom (PKA) energy and the Doppler broadening of reaction cross sections. The former is shown not important due to the threshold energy of atomic displacement and the small value of thermal kinetic energy. For 56Fe, the Doppler broadening has limited influence on DPA cross sections due to high resonance energies. On the other hand, the self-shielding should be considered in the computation of DPA, which depends not only on the cross sections but also the corresponding recoil energies of the PKA. ECCO 1968-group calculations show that the self-shielding corrections on DPA cross sections are necessary for the computation of atomic displacement. In addition, the total DPA rate calculated by self-shielding corrected total cross sections is smaller than the sum of partial components. The recommended method to compute DPA is the calculation of each reaction, especially for coarse energy structures. The total DPA rate in ASTRID reactor inner core is 25 DPA/year, for which the contributions of the elastic scattering, total inelastic scattering, and disappearance reactions are 81.9%, 18.0%, and 0.1%, respectively. The corresponding relative reductions due to the corrections of multi-group cross sections are 11.3%, 5.9%, and 20.5%, respectively.

Application of special relativity in damage calculations

The Displacement per Atom (DPA) is one of the most important factors that influence the mechanical behaviors of materials during and after irradiation. However, the relativistic effect is not accounted for in typical calculations of DPA cross sections and Primary Knock-on Atom (PKA) spectra. The present work shows the computation of relativistic recoil energy. For incident energy lower than 20 MeV, the relativistic corrections are within 1% for neutron scattering reactions of 56Fe and nuclei with mass larger than 56Fe. The corresponding corrections on damage energy (equivalent to DPA with standard DPA metrics) is smaller than 0.5%. However, the relativistic effect is of importance for high incident energies, such as 200 MeV. For example, the relativistic recoil energy of continuum inelastic scattering is 10% higher than the classical one for all nuclei. The consideration of relativistic effect increases the recoil energy for the elastic scattering. Both positive and negative relativistic corrections are possible for both discrete and the inelastic scatterings. A consequent result is the broadening of PKA spectra.

Grain size reduction strategies on Eurofer

One of the options currently taken into account for the realization of the first DEMO reactor is the “water-cooled blanket”. This option implies a minimum irradiation temperature for the blanket material in the range of 280–350 °C. In addition to the DBTT (Ductile to Brittle Transition Temperature) shift due to the DPA (displacement per atom) damage under irradiation, also the issue of the increased embrittlement due to He production must be taken into account. This issue appears even more detrimental and less manageable because the DBBT shift due to the Helium production does not saturate with the dose, as it results from previous works reported in literature. The experimental results and the difference in behaviour between ODS (Oxide Dispersion Strengthened Steels) RAFM (Reduced Activation Ferritic Martensitic) and other FM (Ferritic Martensitic) alloys (EM10, P91) showed that it is possible to improve the resistance to He embrittlement by both intra-granular precipitation of Y-Ti oxides and by decreasing the grain size at the same time. Nevertheless, anyway, the multiplication of the grain boundaries increases the dilution of He on grain surface, delaying the formation of He bubbles on grain boundaries and, therefore, the susceptibility to the He embrittlement. Several grain size reduction strategies have then been investigated on EUROFER both at the austenitization stage, on the PAGS (Prior Austenite Grain Size), and at the tempering stage, on the tempered martensite. The microstructural observations have been carried out by means of SEM (Scanning Electron Microscopy). Also the effect of grain size reduction on the toughness of the material will be taken into account; The DBTTs resulting from impact tests on KLST specimens will be shown. The outcomes of the microstructural observations, as well as the preliminary mechanical characterization (impact tests) will be discussed in this paper.

Characterization of defect clusters in ion-irradiated tungsten by X-Ray diffuse scattering

Defect clusters in ion-irradiated tungsten are characterized by X-ray diffuse scattering. Single crystal tungsten samples are irradiated at 300 K by 0.5–5 MeV copper ions with total doses of 0.2, 0.6, 2, and 5 displacements per atom (DPA). Analysis of the diffuse scattering intensities allows the determination of the concentration of vacancy- and interstitial-type dislocation loops as a function of loop radius for each sample. More vacancies than interstitials are observed in each sample, and the radius of the vacancy loops increases with irradiation dose. For the 0.2DPA sample the average vacancy loop radius is 10.1±3.0Å with a defect density (counting all atoms in the loops) of (10.2±1.5)×10−4 per tungsten atom, increasing to an average loop radius of 16.1±3.1Å at a density of (8.5±1.0)×10−4 per tungsten atom for the 5DPA sample.

The effect of ion irradiation on the electron work function of stainless steel

Stainless steel samples AISI 321 were iradiated with 100 keV H+ and 30 keV Ar+ ions in order to reveal the eventual variation of electron work function, Φ due to high energy radiation. The practical aspect of the research was connected to the change of the steel's corrosion resistance in a nuclear reactor. Work function was measured by the Ambient Kelvin Probe method. Irradiation by Ar+ was found to bring about a decrease whereas that by H+ an increase in the work function. To our knowledge these are the first attempts to reveal the effect of ion irradiation on metal work function. The change, ΔΦ exhibited extrema as a function of displacement per atom (DPA) in both cases. The observations were interpreted in terms of the Lang-Kohn theory which describes Φ as a difference between a surface and a bulk potential. The two types of radiation influence the metal density in the same sense and extent but Ar+ ions affect mainly the near-surface region whereas H+ ions act in the bulk. The opposing signs of the contributions from the two regions account for the opposing effects. Since ΔΦ is a linear function of the variation of the electrode potential, irradiation by Ar+ weakens whereas that by H+ strengthens the alloy's corrosion stability, as far as thermodynamics is concerned.

Damage microstructure evolution of helium ion irradiated SiC under fusion relevant temperatures

In-situ transmission electron microscopy (TEM) with ion irradiation has been used to study the damage microstructure evolution of He ion irradiated 4-H SiC at nuclear fusion relevant temperatures. The SiC samples were irradiated with 20 keV He ions at 25, 400, 800 and 1200 °C to a dose of 5.0 displacements per atom (DPA). At 25 °C, the material fully amorphises at 1.5 DPA and no He bubble nucleation occurs up to the doses studied. At 400 and 800 °C, He bubble nucleation occurs and the material remains crystalline. Bubble nucleation occurs at 2.0 DPA at 400 °C but occurs at only 0.5 DPA at 1200 °C. This is attributed the He atoms de-trapping from vacancies and migrating interstitially to larger He-vacancy clusters at higher temperatures, leading to faster nucleation of observable He bubbles. Helium platelets form at an irradiation temperature of 1200 °C at 0.5 DPA showing a preference for nucleation between the {0001} basal planes.

Thermal analysis at target for neutron generation in fissile assay

A target was designed to produce intense neutrons of 1012 neutrons/sec in order to analyze the fissile content by interrogation in the Lead Slowing Down Spectrometer (LSDS) system. A high energy and current electron linear accelerator was used in neutron production by bremsstrahlung radiation. However, the interaction by electron and radiation in the target generates large amounts of heat. Several coolants were simulated in the target and the heat removal capacity was performed. The thermal property of coolant in the target was calculated using the CFX code. The heat distribution by the cooling method was simulated and the forced convection method was used to calculate the temperature distribution on the different coolants and flow rate. Helium shows a good cooling property, but it needs high flow rate with high pressure. Air may be a good candidate coolant if the limit of neutron activation is satisfied in the circulating system. When the coolants were used in the target, each metal plate of the target was kept under the melting temperature. For the estimation of the coolant performance, the neutron influence by the coolant was examined as well. The material property was also evaluated for various target cases. The displacement per atom (DPA) analysis of the target was performed for a life time operation of the LSDS system. For the target container material (HT-9), the optimum thickness was obtained from the sensitivity simulation.

Improved resource utilization by advanced burner reactors with breed-and-burn blankets

Previous studies assessed the feasibility of designing Sodium-cooled Fast Reactors (SFR) in a novel Seed-and-Blanket (S&B) configuration in which a significant fraction of the core power is generated by a radial metallic thorium-fueled blanket that operates in the Breed-and-Burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 Displacements per Atom (DPA). This paper investigates a number of blanket fuel options other than metallic thorium fuel, including thorium dioxide fuel, thorium hydride fuel, thorium nitride Fully Ceramic Encapsulated (FCM) fuel, and Pressurized Water Reactor (PWR) Used Nuclear Fuel (UNF) after limited reprocessing. Since the neutron spectrum of the oxide fueled blanket is softer than that of the reference metallic thorium fueled blanket, it can discharge its fuel at an average burnup of around 110 MWd/kg versus 65 MWd/kg of the metallic thorium. The two thorium hydride fueled blankets feature an even softer neutron spectrum than the oxide fueled blanket and, therefore, a higher average discharge burnup — 192 MWd/kg and 245 MWd/kg when the H-to-Th ratio is, respectively, 0.5 and 2.0. The FCM fuel is composed of SiC matrix and cladding that is expected to self-anneal the neutrons induced damage. With the assumption that the FCM fuel will indeed be proven not limited by radiation damage, its discharge burnup could be up to 482 MWd/kg. As a result, the corresponding thorium utilization is the highest of all options examined -- close to 80 times the utilization of natural uranium in present Light Water Reactors (LWRs). The amount of Trans-Th isotopes accumulated in the thorium blanket per unit of generated electricity decreases as the blanket fuel is discharged at a higher burnup. Therefore, a higher discharge blanket burnup results in lower long-term radioactivity and radiotoxicity. It is also found that the 232U/233U ratio in the discharged thorium fuel is over 3 times higher in the thorium hydride than in the other blankets thus improving the thorium-hydride blanket's proliferation resistance. Softening of the blanket spectrum leads to softening of the seed spectrum region interfacing with the blanket and this enables to discharge the seed fuel at a higher average burnup. The higher discharge burnup of the seed fuel reduces the required reprocessing and fuel fabrication capacities per unit of generated electricity. The cores having softer neutron spectrum blankets also feature less positive feedback to sodium voiding and much more negative Doppler coefficient.When PWR UNF is used after limited reprocessing to fuel the blanket, the subcritical blanket driven by the excess neutrons from the seed can discharge this fuel at an average burnup of ∼120 MWd/kg. Combined with the burnup in the first stage PWRs, this two-stage energy system can achieve an accumulated uranium burnup of ∼170 MWd/kg. This increases the fuel utilization of natural uranium by a factor of three relative to once-through PWRs while at the same time reducing the amount of high-level waste that needs to be disposed of per unit of generated electricity.Another scenario examined in this paper is a 3-stage closed energy system: TRU extracted from Stage-1 PWRs used fuel is charged to the seed of Stage-2 S&B cores while the Trans-Th (mainly 233U) extracted from Stage-2 S&B blanket discharged fuel is used as the fissile feed for Stage-3 PWRs. It is estimated that 1GWe of S&B SFRs in such a 3-Stage energy system can support 3.2 GWe of PWRs versus ∼1.7 GWe of PWRs that can be supported by 1 GWe of CR = 0.5 Advanced Burner Reactor (ABR).

Implementing displacement damage calculations for electrons and gamma rays in the Particle and Heavy-Ion Transport code System

Abstract In this study, the Monte Carlo displacement damage calculation method in the Particle and Heavy-Ion Transport code System (PHITS) was improved to calculate displacements per atom (DPA) values due to irradiation by electrons (or positrons) and gamma rays. For the damage due to electrons and gamma rays, PHITS simulates electromagnetic cascades using the Electron Gamma Shower version 5 (EGS5) algorithm and calculates DPA values using the recoil energies and the McKinley–Feshbach cross section. A comparison of DPA values calculated by PHITS and the Monte Carlo assisted Classical Method (MCCM) reveals that they were in good agreement for gamma-ray irradiations of silicon and iron at energies that were less than 10 MeV. Above 10 MeV, PHITS can calculate DPA values not only for electrons but also for charged particles produced by photonuclear reactions. In DPA depth distributions under electron and gamma-ray irradiations, build-up effects can be observed near the target’s surface. For irradiation of 90-cm-thick carbon by protons with energies of more than 30 GeV, the ratio of the secondary electron DPA values to the total DPA values is more than 10% and increases with an increase in incident energy. In summary, PHITS can calculate DPA values for all particles and materials over a wide energy range between 1 keV and 1 TeV for electrons, gamma rays, and charged particles and between 10−5 eV and 1 TeV for neutrons.

Investigation of DPA in the reactor pressure vessel of VVER-1000/V320

The most important ageing effect on the reactor pressure vessel (RPV) is radiation embrittlement, which is mainly caused by fast neutrons during operation lifetime of nuclear reactors. The aim of this study was to investigate the DPA (displacement per atom) rate, an important parameter describing radiation damage to the RPV, and identify the position of the maximum DPA rate in the RPV of the VVER-1000/V320 reactor using the Monte Carlo code MCNP5. To reduce statistical errors in the MCNP5 simulation, the weight window technique was applied to non-repeated structures outside the reactor core. The results showed the distribution of the DPA rate in the RPV and the maximum DPA rate was found to be at the first millimeters of the RPV. Consequently, these calculations could be useful for assessment of radiation damage to the RPV of VVER reactors.

Displacement damage cross section and mechanical properties calculation of an Es-Salam research reactor aluminum vessel

Nuclear facility aging is one of the biggest problems encountered in nuclear engineering. Radiation damage is among one of the aging causes. This kind of damage is an important factor of mechanical properties deterioration. The interest of this study is on the Es-Salam research reactor aluminum vessel aging due to neutron radiation. Monte Carlo (MC) simulations were performed by MCNP6 and SRIM codes to estimate the defects created by neutrons in the vessel. MC simulations by MCNP6 have been performed to determine the distribution of neutron fluence and primary knock-on atom (PKA) creation. Considering our boundary conditions of the calculations, the helium and hydrogen gas production in the model at a normalized total neutron flux of 6.62 × 1012 n/cm2 s were determined to be 2.86 × 108 and 1.33 × 109 atoms/cm3 s, respectively. The SRIM code was used for the simulation of defects creation (vacancies, voids) in the aluminum alloy of the Es-Salam vessel (EsAl) by helium and hydrogen with an approximate energy of 11 MeV each. The coupling between the two codes is based upon post-processing of the particle track (PTRAC) output file generated by the MCNP6. A small program based on the MatLab language is performed to condition the output file MCNP6 in the format of a SRIM input file. The concentration of silicon was determined for the vessel by the calculation of the total rate of 27Al(n,γ)28Si reaction. The DPA (displacement per atom) was calculated in SRIM according to R.E. Stoller recommendations; the calculated value is 0.02 at a fast neutron fluence 1.89 × 1019 n/cm2. RCC-MRx standard for 6061-T6 aluminum was used for the simulation of the evolution of mechanical properties for high fluence. The calculated values of nuclear parameters and DPA obtained were in agreement with the experimental results from the Oak Ridge High Flux Isotope Reactor (HFIR) reported by Farrell and coworkers.

Fuel cycle analysis of Advanced Burner Reactor with breed-and-burn thorium blanket

The Seed-and-Blanket (S&B) Sodium-cooled Fast Reactor (SFR) core concept was proposed for generating a significant fraction of the core power from thorium fueled breed-and-burn (B&B) blankets without exceeding the presently verified radiation damage constraint of 200 Displacement per Atom (DPA). To make beneficial use of the excess neutrons from fast reactors, the S&B core is designed to have an elongated TRU transmuting (or “TRU burner”) seed from which over 20% of the fission neutrons leak into a subcritical thorium blanket that radially surrounds the seed. The seed fuel is recycled while the blanket operates in a once-through breed-and-burn (B&B) mode. The objective of this paper is to compare the fuel cycle performance of the S&B reactor against an Advanced Burner Reactor (ABR) and a conventional Pressurized Water Reactor (PWR). For the fast reactors (SFR: ABR and S&B) the fuel cycle performance is evaluated based on a 2-stage PWR-SFR energy system while the reference nuclear system is made of once-through PWRs.It was found that relative to the ABR, the S&B core has a lower fuel cycle cost, higher capacity factor, and comparable short-term radioactivity. The discharged seed fuel from the S&B core features lower fissile Pu-to-Pu ratio, higher 238Pu-to-Pu ratio, higher specific plutonium decay heat, higher spontaneous fission rate, and lower overall material attractiveness for weapon use. Due to the significant amount of 233U discharged from the breed-and-burn thorium fueled blankets, the S&B core has much higher long-term radioactivity and radiotoxicity. Since the thorium fueled blanket operates in the breed-and-burn mode and requires no fuel reprocessing, the discharged blanket fuel is unattractive for weapons application.Compared with a PWR, the S&B core has a lower fuel cycle cost, much lower short-term radioactivity and radiotoxicity but higher long-term values, and higher proliferation resistance for the discharged plutonium. The natural uranium utilization of the 2-stage PWR-S&B system is approximately 60% higher than that of present PWRs; it is few percent higher than that of the 2-stage PWR-ABR system. Approximately 7% of the thorium fed to the blanket is converted into energy, which makes the thorium fuel utilization approximately 12 times the utilization of natural uranium in PWRs.A comprehensive fuel cycle evaluation performed with the methodology developed by the recent U.S Department of Energy’s Nuclear Fuel Cycle Evaluation and Screening campaign concludes that the PWR-S&B system has similar fuel cycle performance characteristics as the PWR-ABR system. The S&B concept may potentially feature improved economics and resource utilization relative to the ABR.

Neutronics performance and activation calculation of dense tungsten granular target for China-ADS

Abstract Spallation target, which constitutes the physical and functional interface between the high power accelerator and the subcritical core, is one of the most important components in Accelerator Driven Subcritical System (ADS). In this paper, we investigated the neutronics performance, the radiation damage and the activation of dense tungsten granular flow spallation target by using the Monte Carlo programs GMT and FLUKA at the proton energy of 250 MeV with a beam current of 10 mA . First, the leaking neutron yield, leaking neutron energy spectrum and laterally leaking neutron distribution at several time nodes and with different target parameters are explored. After that, the displacement per atom (DPA) and the helium/hydrogen production for tungsten grains and structural materials with stainless steel 316L are estimated. Finally, the radioactivity, residual dose rate and afterheat of granular target are presented. Results indicate that granule diameter below 1 cm and the beam profile diameter have negligible impact on neutronics performance, while the target diameter and volume fraction of grain have notable influence. The maximum DPA for target vessel (beam tube) is about 1.0 (1.6) DPA/year in bare target, and increased to 2.6 (2.8) DPA/year in fission environment. Average DPA for tungsten grains is relatively low. The decline rate of radioactivity and afterheat with cooling time grows with the decrease of the irradiation time.

Evaluation of radiation damage in belt-line region of VVER-1000 nuclear reactor pressure vessel

Regarding irradiation damage, the most sensitive region in the reactor pressure vessel (RPV) is the area adjacent to the reactor core; the so called belt-line region. Belt-lines and their related heat affected zones (HAZs) in the belt-line region are constantly under neutron and gamma irradiation and have a higher potential to various flaws. The microscopic defects produced in materials due to irradiation are referred to as radiation damage; the units of displacements per atom (DPA) are used routinely to characterize the extent of such damage. In this research, for a typical VVER-1000 (V446) RPV, the areas of the highest neutron radiation flux (belt-line) is first identified by using a Monte Carlo neutron transport code. This is used to calculate and validate the neutron flux and fluence at the belt-line. Employing the PTRAC output from MCNP and the SRIM Monte Carlo code, radiation damage in the RPV is simulated and assessed in terms of DPA. The integrated damage for a full effective year is evaluated to be order of 5E-04 displacements per cm thickness of RPV belt-line metal.

COMPARISON OF DPA AND HELIUM PRODUCTION IN CANDIDATE FUEL CLADDING MATERIALS FOR THE CANADIAN SCWR

Harsh conditions associated with high-temperature, high-pressure supercritical-water coolant in the Canadian supercritical water-cooled reactor (SCWR) necessitate the use of stainless steel or nickel-based alloys as fuel cladding material rather than more conventional zirconium-based alloys. In-core neutron irradiation is expected to produce helium in nickel-bearing alloys through the 2-step 58Ni(n,γ)59Ni(n,α)56Fe reaction. Increases in displacements per atom (DPA) associated with this reaction will also occur. It is important to quantify neutron-irradiation induced helium production and DPA to assess the consequences for fuel sheath material integrity. In this paper, neutron-flux induced helium production and displacement damage are estimated from lattice physics calculations for 5 candidate SCWR fuel clad materials, alloys 214, 625, 800 H, 347, and zirconium-modified 310 stainless steel, and compared for 2 fuel options, plutonium-thorium dioxide, (Pu–Th)O2, and low enriched uranium-thorirum dioxide, (LE...

Delocalized Plastic Flow in Proton-Irradiated Monolithic Metallic Glasses.

Creating new materials with novel properties through structural modification is the Holy Grail of materials science. The range of targetable structures for amplification of mechanical properties in metallic glasses would include types of atomic short range orders at the smallest scale through compositions or morphologies of phases in composites. Even though the usefulness of the latter approach has been successfully demonstrated in the past decades, the feasibility of the former has been incompletely proved with only marginal property improvements reported within experimentally-accessible atomic-level structural changes. Here, we report the significant enhancement of deformability in Zr-based monolithic metallic glass only through the atomic disordering by proton irradiation without altering any other structural traits. Metallic glass nanopillars that originally failed catastrophically without any notable plasticity become capable of attaining more than 30% uniaxial plastic strain accommodated by homogeneous deformation when irradiated to ~1 displacement per atom (DPA). We discuss the atomistic origin of this improved plasticity in terms of density and spatial distributions of icosahedral short range order influenced by irradiation.