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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Effects of helium on irradiation response of reduced-activation ferritic-martensitic steels: Using nickel isotopes to simulate fusion neutron response

Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58 Ni- and 60 Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58 Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60 Ni-doped samples. The greater helium production of 58 Ni did not show pronounced reductions in ductility of the samples.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Microstructures and mechanical properties of a modified 9Cr ferritic-martensitic steel in the as-built condition after additive manufacturing

A newly developed nano-structured high-Mn 9Cr ferritic-martensitic (FM) steel designed for additive manufacturing (Additive-manufactured Nanostructured Alloy, ANA) was fabricated via the direct energy deposition method. The as-printed ANA2 showed tensile properties (strength and elongation) and Charpy impact toughness (upper-shelf energy and ductile-brittle transition temperature) comparable to oxide-dispersion-strengthened (ODS) Eurofer and PM2000. Compared with wrought P91, the as-built ANA2 showed significantly higher yield strength but inferior Charpy impact toughness and ductility. Microstructural characterization revealed uniformly distributed sub-micron pores, an extremely high dislocation density, and a high density of ultrafine nano-structured precipitates in the matrix. The microstructures indicated the dislocations as the dominant strengthening element and the ultrafine precipitates as the primary contribution to estimated irradiation sink density. This work demonstrates the ability to exploit the unique consolidation properties of additive manufacturing to fabricate steels which marry the benefits of both ODS alloys and conventionally processed FM steels.

36 MATERIALS SCIENCE↗
DOE OSTI2021

The Role of Co-injected Helium on Swelling and Cavity Evolution at High Damage Levels in Ferritic-Martensitic Steels

The influence of helium co-injection at rates from 0 to 4 appm He/dpa on swelling in ferritic-martensitic alloys T91 and HT9 was explored. Irradiations with 5.0 MeV Fe++ ions and degraded He++ ions were performed at 445°C up to damage levels of 150 dpa and helium co-injection rates of 0, 0.02, 0.2 and 4 appm He/dpa in T91, and at 460°C to a damage level of 188 dpa and helium co-injection rates of 0, 0.06 and 4 appm He/dpa in HT9. Helium was observed to enhance cavity nucleation at low damage levels, resulting in the maximum swelling at the highest helium co-injection rate. As the damage level was increased, the helium content at which swelling was maximized shifted to lower helium concentrations, ultimately resulting in the highest swelling occurring with zero helium by 150 dpa. Finally, this behavior was due to the helium-stabilized bubble microstructure that increased the cavity sink strength and reduced both cavity growth rate and swelling relative to the helium-free condition.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Under the microscope: Reduced activation ferritic martensitic steel Eurofer-97 following ion-Irradiation and high-temperature high-pressure water exposure

Here, this study is designed to characterise the microstructural behaviour of Eurofer-97 steel under ion irradiation and subsequent exposure to high-temperature high-pressure (HTHP) water. Eurofer-97, a ferritic-martensitic steel, has been developed to withstand the conditions of fusion reactors in the locations in contact with coolant with an elevated level of neutron flux, such as the breeder-wall blanket. The material has been studied after self-ion irradiation (using Fe ions) simulating the microstructural effects of neutron irradiation limited to the subsurface layer. The corrosion properties of the Eurofer-97 steel were studied by exposure to HTHP water up to 331 °C. Advanced microstructural characterisation using scanning, transmission electron and focused ion beam microscopy was performed on the as-received microstructure and after ion irradiation. This was then characterised after exposure for 240 h in high-temperature water. Eurofer-97 had a dense, columnar Cr-rich inner oxide, followed by a Fe-rich outer oxide layer. In the irradiated condition the grain structure and oxide itself was less ordered. No appreciable difference in oxide thickness was identified between the irradiated and unirradiated specimens after this short exposure time.

36 MATERIALS SCIENCE↗
DOE OSTI2025

High-temperature nanoindentation creep studies on castable and sintered nanostructured low-activation ferritic-martensitic alloys

In this article, we present the creep characteristics of two reduced activation ferritic-martensitic steels of identical starting compositions formed by different fabrication routes: a nanostructured ferritic alloy commonly referred to as a castable nanostructured alloy (CNA) and a sintered nanostructured alloy (SNA) variant. Through a series of nanoindentation experiments spanning a temperature range of 25 °C to 650 °C, with a maximum load of 100 mN, we find creep behaviors in the cast and sintered materials to be remarkably similar. The creep stress exponent (n) for CNA and SNA were found to be in the range of 8–35 and the activation volume was ∼14–42b 3 , underscoring a dominance of dislocation-mediated mechanisms in both alloys. Notably, we observed a decline in the creep stress exponent with increasing temperature, attributable to the heightened influence of thermally activated dislocations. This phenomenon suggests a potential transition in the deformation mechanism towards a thermally activated dislocation climb process, significantly impacting the observed creep behavior.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Comparison of void swelling of ferritic-martensitic and ferritic HT9 alloys after high-dose self-ion irradiation

In this work, ferritic-martensitic (F/M) HT9 along with fully ferritic HT9 specimens were irradiated using self-ions to 600 peak displacement-per-atom (dpa) at 450 °C. To investigate and minimize the carbon contamination effect on void swelling, two identical F/M HT9 specimens were prepared: one without a coating and the other one with a tungsten and silver coating on the surface prior to irradiation. The specimen without a coating showed carbide formation in the matrix, resulting from carbon contamination during irradiation, with no voids observed. The coated sample did not form matrix carbides, and voids were observed within the irradiated region. The same coating was applied to the ferritic HT9, and void swelling results of both the F/M HT9 and the ferritic HT9 alloys were compared after irradiation. The F/M HT9 exhibited a maximum of 2.6% swelling, while the ferritic HT9 showed a maximum of 0.8% swelling at a 400–500 nm depth, equivalent to 334 local dpa. The results were correlated with the instability of carbides in the ferritic HT9 under irradiation. Carbides in the F/M HT9 and the ferritic HT9 are both M23C6, but the matrix carbides in the ferritic HT9 were less stable under the irradiation, compared to the ones on the grain boundaries in the F/M HT9. The study shows the impact of chemistry on the void swelling which overpowers the microstructural factors such as grain size and phase.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Investigation of the fatigue crack behavior of 304 stainless steels using synchrotron X-ray tomography and diffraction: Influence of the martensite fraction and role of inclusions

Here, the effect of fatigue on the microstructure of four-point bend specimens of three variations of 304 stainless steels (Commercial 304, 304H, and 304 L) was investigated using synchrotron x-ray tomography and diffraction. X-ray tomography revealed the formation of the fatigue-induced microvoids and crack while the diffraction data was used to quantify the amount of deformation-induced martensite found after fatigue in all samples. Transmission electron microscopy evidenced the role of the precipitates/inclusions on the microvoid formation. It was found to depend on their chemical nature. The shape of the precipitates/inclusions was also found to have an effect on the microvoid shape.

304 stainless steel↗
DOE OSTI2020

Microstructural stability of tantalum-alloyed ferritic-martensitic steel with neutron irradiation to 7.4 dpa at ~490 °C

Specimens of a reduced-activation ferritic-martensitic steel, called CNA1, were irradiated to 7.4 displacements per atom at ~490 °C in the High Flux Isotope Reactor. Vickers hardness and tensile properties revealed consistent softening that is attributed to some recovery during irradiation, with reduced laths and dislocation density. Detailed microstructural characterization by transmission electron microscopy and energy dispersive x-ray spectroscopy showed radiation-induced dislocation loops, few cavities, and slight dissolution of M 23 C 6 and MX precipitates. Partial amorphization was only observed in M 23 C 6 precipitated in matrix. Transmutation somewhat destabilized M 23 C 6 and MX, with a stronger effect on the later one by noticeable depletion of Ta with enriched W, which is consistent with the transmutation calculations. The microstructural evolution and associated strength changes were discussed with the assistance of computational thermodynamics.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A machine learning aided interpretable model for rupture strength prediction in Fe-based martensitic and austenitic alloys

Abstract The class of 9–12% Cr ferritic-martensitic alloys (FMA) and austenitic stainless steels have received considerable attention due to their numerous applications in high temperature power generation industries. To design high strength steels with prolonged service life requires a thorough understanding of the long-term properties, e.g., creep rupture strength, rupture life, etc., as a function of the chemical composition and processing parameters that govern the microstructural characteristics. In this article, the creep rupture strength of both 9–12% Cr FMA and austenitic stainless steel has been parameterized using curated experimental datasets with a gradient boosting machine. The trained model has been cross validated against unseen test data and achieved high predictive performance in terms of correlation coefficient ( $$R^{2} > 0.98 $$ R 2 > 0.98 for 9–12% Cr FMA and $$R^{2} > 0.95 $$ R 2 > 0.95 for austenitic stainless steel) thus bypassing the need for additional comprehensive tensile test campaigns or physical theoretical calculations. Furthermore, the feature importance has been computed using the Shapley value analysis to understand the complex interplay of different features.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A room temperature exchange bias effect caused by the coexisting martensitic phase structures in Ni50Mn38Sb12−xGax polycrystalline Heusler alloys

The exchange bias effect is the physical cornerstone of applications, such as spin valves, ultra-high-density data storage, and magnetic tunnel junctions. This work studied the room temperature exchange bias effect by constructing a Ni50Mn38Sb12−xGax alloy system with coexisting martensitic phase structures. The study found that the exchange bias effect shows a non-monotonic change with the variation of Ga composition at 300 K, and an obvious room temperature exchange bias effect appears in the alloys with coexisting phase structures of 4O and L10, which is due to the strong exchange coupling between ferromagnetic and antiferromagnetic. Further research on the exchange bias effect and temperature shows that the blocking temperature is 420 K, and the exchange bias can stably exist in a temperature range of ∼200 K around room temperature. This work provides a method to engineer exchange bias effects at room temperature.

Physics↗
DOE OSTI2021

Martensitic Transformation in Fe x Mn 80-x Co 10 Cr 10 High-Entropy Alloy

High-entropy alloys and even medium-entropy alloys are an intriguing class of materials in that structure and property relations can be controlled via alloying and chemical disorder over wide ranges in the composition space. Employing density-functional theory combined with the coherent-potential approximation to average over all chemical configurations, we tune free energies between face-centered-cubic and hexagonal-close-packed phases in Fe x Mn 80 - x Co 10 Cr 10 systems. Within Fe-Mn-based alloys, we show that the martensitic transformation and chemical short-range order directly correlate with the face-centered-cubic and hexagonal-close-packed energy difference and stacking-fault energies, which are in quantitative agreement with recent observation of two phase region (face-centered cubic and hexagonal closed pack) in a polycrystalline high-entropy alloy sample at x = 40 at.%. Our predictions are further confirmed by single-crystal measurements on a x = 40 at.% using transmission-electron microscopy, selective-area diffraction, and electron-backscattered-diffraction mapping. The results herein offer an understanding of transformation-induced or twinning-induced plasticity in this class of high-entropy alloys and a design guide for controlling the physics at the electronic level.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2020

Second Annual Progress Report on Correlation Between Microstructure and Mechanical Properties of Neutron-Irradiated Ferritic-Martensitic and Austenitic Steels

Ferritic-martensitic steels G92-2b (an optimized Grade 92 heat), NF616 and T91, and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) and the Advanced Test Reactor (ATR) of Idaho National Laboratory (INL). Selected G92-2b samples were irradiated up to 14.66 dpa in the HFIR at two temperature ranges: 400–496.7°C and 683.3– 720°C. NF616 and T91 were irradiated in the ATR up to 8.16 dpa with the irradiation temperatures ranged from 241°C to 447.5°C. Alloy 800H and 800H-TMP samples were irradiated in both the HFIR and the ATR. Selected 800H and 800H-TMP samples had HFIR irradiation to 1.28 dpa at 580°C and ATR irradiation up to 9.12 dpa at 359°C to 431°C. Vickers hardness measurements, fractography, and microstructural characterization were performed on the selected samples in the Low Activation Materials Design and Analysis (LAMDA) laboratory. Radiation-hardening of G92-2b was observed at the lower doses and lower irradiation temperatures (400- 496.7°C), with GB03 (0.52 dpa at 400°C) and GB04 (7.44 dpa at ~490°C) showing ~12% and ~8% hardening, respectively. Softening by ~14% was observed for GB05 (14.66 dpa at 496.7°C). Radiationsoftening of G92-2b was more prevalent at the higher irradiation temperatures (683.3-~720°C), with GB10 (0.46 dpa at 683.3°C), GB11 (7.44 dpa at ~720°C), and GB12 (14.63 dpa at ~720°C) showing ~8%, ~8%, and ~40% softening, respectively. Radiation-hardening of NF616 and T91 was observed with the hardness increased by ~37% to ~65% depending on the irradiation doses and irradiation temperatures. Within the studied irradiation conditions of NF616 and T91, samples with a higher dose had a larger hardness after irradiation. All the tested alloy 800H and 800H-TMP samples in this work showed radiation-hardening by ~96±7% to ~152±10%. Alloy 800H-TMP tended to have slightly smaller radiation-hardening than alloy 800H. The fractography results of G92-2b sample GB03, GB10, and GB11, together with the previously characterized fractography of GB04, GB05, and GB12, indicated that the ductility of G92-2b was maintained up to 14.66 dpa at the lower irradiation temperatures of 400-496.7°C, while some loss of ductility (less necking) was observed for higher doses at the higher irradiation temperatures of 683.3-720°C. This agrees with the previously reported tensile test results of G92-2b, where the elongation of G92-2b was reduced at higher doses at the higher irradiation temperatures. Dimple sizes increased at higher doses, which are more evident at the higher irradiation temperatures of 683.3-~720°C. The fractography of NF616 sample D2 (2.96 dpa at 291.5°C), D4 (5.91 dpa at 359°C), and D6 (8.16 dpa at 431°C) indicated loss of ductility with negligible necking for sample D2, while ductile failure for samples D4 and D6. Fractography of alloy 800H and 800H-TMP samples in various irradiation conditions showed ductile failure with obvious necking. Dimples were observed, with some of them containing large Ti-rich particles, in all the 800H/800H-TMP samples. Electron backscatter diffraction characterization of GB12 indicated the recovery of the lath structure, which was generally replaced by an equiaxed grain structure. Transmission electron microscopy (TEM) characterization showed the presence of frequent M 23 C 6 (M = primarily Cr), MX (M = primarily V), spherical Nb(C,N) precipitates, and occasional Laves phase precipitates in the G92-2b samples. MX precipitates with sizes of 20-30 nm were observed at boundaries of smaller grains, indicating the pinning effect of the V-rich precipitates. The lath structure recovery was more evident at the higher irradiation temperatures (683.3-~720°C), with decreased densities of line dislocations and M23C6 precipitates. The irradiated T91 (TA04) showed the growth of M23C6 precipitates to 101 ± 40 nm from the initial 68 ± 22 nm in the unirradiated condition. Dislocation loops of both {100} and {111} types were present in TA04. TEM characterization was also performed on the irradiated 800H (N4, N5, N6, and AR2) and 800H-TMP (P4, P5, P6, and HG1). Accumulation of M 23 C 6 precipitates at grain boundaries was observed in all the xii 800H/800H-TMP samples, and the presence of Ti(C,N) precipitates at grain boundaries and in the matrix was observed in the irradiated 800H-TMP. Some Ti(C,N) precipitates are embedded in the M 23 C 6 precipitates, maintaining specific orientation relationships between the precipitates and between the precipitate and matrix. In addition, nanoscale Si-rich clusters were observed in the matrix of all the 800H/800H-TMP samples, with EDS Si maps tending to have a lower contrast in 800H-TMP samples. Atom probe tomography was conducted on the same samples, supported by a Rapid Turnaround Examination project under Nuclear Science User Facilities. The results are being analyzed to be integrated with the TEM results for a confident description of the γ’ precipitates. Dislocation loops also formed in all the 800H/800H-TMP samples. The density and the average size of dislocation loops were quantified to be in the order of 10 22 – 10 23 m -3 and 11.7 – 15.9 nm, respectively, in the ATR-irradiated 800H/800H-TMP samples. The loop densities in irradiated 800H were higher than that in irradiated 800H-TMP under the same irradiation conditions. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.

36 MATERIALS SCIENCE↗
DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Modeling of Stress and Temperature Effects on Creep of Reduced Activation Ferritic-Martensitic Steel Alloy F82H (Tertiary Creep Modeling of RAFM Steel)

A Bayesian optimization procedure is presented for calibrating a multi-mechanism micromechanical model for creep to experimental data of F82H steel. Reduced activation ferritic martensitic (RAFM) steels based on are the most promising candidates for some fusion reactor structures. Although there are indications that RAFM steel could be viable for fusion applications at temperatures up to 600 °C, the maximum operating temperature will be determined by the creep properties of the structural material and the breeder material compatibility with the structural material. Due to the relative paucity of available creep data on F82H steel compared to other alloys such as Grade 91 steel, micromechanical models are sought for simulating creep based on relevant deformation mechanisms. As a point of departure, this work recalibrates a model form that was previously proposed for Grade 91 steel to match creep curves for F82H steel. Due to the large number of parameters (9) and cost of the nonlinear simulations, an automated approach for tuning the parameters is pursued using a recently developed Bayesian optimization for functional output (BOFO) framework [1]. Incorporating extensions such as batch sequencing and weighted experimental load cases into BOFO, a reasonably small error between experimental and simulated creep curves at two load levels is achieved in a reasonable number of iterations. Validation with an additional creep curve provides confidence in the fitted parameters obtained from the automated calibration procedure to describe the creep behavior of F82H steel at 600 °C. The model is further extended using a temperature dependent scaling law approach to simulate creep response between 550 °C and 650 °C. The efficacy of this extension is compared with the previously used scaling law approach for Grade 91 steel.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Creep Resistant Martensitic Steels for Operation at High-Temperatures in Power Generation Applications

Increasing the temperature capabilities of ferritic/martensitic 9-12% Cr steels can help in increasing the operating temperature of land-based turbines and minimize the use of expensive high-temperature alloys in the hot section. A creep resistant martensitic steel, JMP, was developed with the potential to operate at or above 650°C. The design of the alloys originated from computational modeling for phase stability and precipitate strengthening using fifteen constituent elements. Cobalt was used for increased solid solution strengthening, Si for oxidation resistance and different W and Mo concentrations for matrix strength and stability. The JMP steels showed increases in creep life compared to MARBN/SAVE12AD at 650°C for testing at various stresses between 138 MPa and 207 MPa. On a Larson-Miller plot, the performance of the JMP steels surpasses that of state-of-the-art MARBN steel. Approximately 21 years of cumulative creep data are reported for the JMP steels which encompasses various compositions. The relationships between composition-microstructure-creep properties are discussed including characterization of microstructures after >20,000 hours in creep.

Detrois, Martin↗
DOE OSTI2025

Primary and Low-Strain Creep Models for 9Cr Tempered Martensitic Steels Including the Effects of Irradiation Softening and High-Helium Re-Hardening

Primary and low-strain creep represents a very important integrity challenge to large, complex structures, like fusion reactors. Here, we develop a predictive empirical primary creep model for 9Cr tempered martensitic steels (TMS), relating the applied stress (σ) to strain (ε), time (t) and temperature (T). The most accurate model is based on the applied σ normalized by the steel’s T-dependent ultimate tensile stress (σo), σ/σo(T). The model, fit to 17 heats of 9Cr TMS, yielded a σ root mean square error (RMSE) of ≈±11 MPa. Notably, the model also provides robust predictions for all the other TMS, when calibrated only by the fusion candidate Eurofer97 database. The model was extended to explore two possible effects of neutron irradiation, which produces both displacements per atom (dpa) and helium (He in atomic parts per million, appm) damage. These effects, which have not been previously considered, include: (a) softening, as a function of dpa, at T > ≈400–450 °C, in low-He fission environments (<1 He/dpa); and (b) subsequent re-hardening in high-He (≥10 He/dpa) fusion first-wall environments. The irradiation effect models predict (a) accelerated primary creep due to irradiation softening; and (b) fully arrested creep due to high-He re-hardening.

Alam, Md Ershadul (ORCID:0000000345968026)↗
DOE OSTI2025

Comparison of Ion and Neutron Irradiations to 3 dpa at 500C in Ferritic-Martensitic Alloys

The growing global demand for energy will increasingly call upon advanced nuclear fission reactors to supply safe and reliable electricity. The structural and fuel cladding components of these reactors will be subject to extreme conditions of irradiation damage up to several hundred displacements per atom (dpa) at temperatures as high as 700°C. Ferritic-martensitic (F-M) steels are leading candidates for these challenging conditions due to their strength and dimensional stability under irradiation. In order to accelerate the process for evaluating F-M alloys, charged particles are increasingly being used to emulate neutron irradiations. Charged particle irradiations allow the possibility of conducting irradiation experiments within a shorter time period (i.e. at a rate up to 4 orders of magnitude faster) and with minimal radioactivation of the material, enabling lower cost and faster turnaround of post irradiation examination and analysis. However, the irradiation dose rate, damage cascade morphologies, and irradiation damage depth profiles all differ widely between protons, heavier ions, and neutrons. Currently, there is limited understanding of the significance of these physical differences and how they manifest in the irradiated microstructure and mechanical properties of F-M steels. The objective of this study is to evaluate charged particles as a surrogate for neutron irradiations in F-M alloys by assessing common irradiation conditions using Fe++ ions, protons, and neutrons. Keeping the temperature and dose consistent enables isolation of the effects of each irradiating particle and their respective dose rates and cascade morphologies.

Swenson, M.J.↗