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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Monte-Carlo Simulations of the Nuclear Energy Deposition Inside the CARMEN-1P Differential Calorimeter Irradiated into OSIRIS Reactor

The nuclear heating measurements in Material Testing Reactors (MTRs) are crucial for the study of nuclear materials and fuels under irradiation. The reference measurements of this nuclear heating are especially performed by a differential calorimeter including a graphite sample material. These measurements are then used for other experimental conditions in order to predict the nuclear heating and thermal conditions induced in the irradiation devices. Nuclear heating is a great deal of interest at the moment as the measurement of such heating is an important issue for MTRs reactors. This need is especially generated by the new Jules Horowitz Reactor (JHR), under construction at CEA/Cadarache 'French Alternative Energies and Atomic Energy Commission'. This new reactor, that will be operational in late 2019, is a new facility for the nuclear research on materials and fuels. Indeed the expected nuclear heating rate is about 20 W/g for nominal capacity of 100 MW. The present Monte Carlo calculation works belong to the IN-CORE (Instrumentation for Nuclear radiation and Calorimetry On line in Reactor): a joint research program between the CEA and Aix- Marseille University in 2009. One scientific aim of this program is to design and develop a multi-sensors device, called CARMEN, dedicated to the measurements of main physical parameters simultaneously encountered inside JHR's experimental channels (core and reflector) such as neutron fluxes, photon fluxes, temperature, and nuclear heating. A first prototype was already developed. This prototype includes two mock-ups dedicated respectively to neutronic measurements (CARMEN-1N) and to photonic measurements (CARMEN-1P) with in particular a specific differential calorimeter. Two irradiation campaigns were performed successfully in the periphery of OSIRIS reactor (a MTR located at Saclay, France) in 2012 for nuclear heating levels up to 2 W/g. First Monte Carlo calculations reduced to the graphite sample of the calorimeter were carried out. A preliminary analysis shows that the numerical results overestimate the measurements by about 20 %. A new approach has been developed in order to estimate the nuclear heating by two methods (energy deposition or KERMA) by considering the whole complete geometry of the sensor. This new approach will contribute to the interpretation of the irradiation campaign and will be useful to improve the out-of-pile calibration procedure of the sensor and its thermal response during irradiations. The aim of this paper is to present simulations made by using MCNP5 Monte-Carlo transport code (using ENDF/B-VI nuclear data library) for the nuclear heating inside the different parts of the calorimeter (head, rod and base). Calculations into two steps will be realized. We will use as an input source in the model new spectra (neutrons, prompt-photons and delayed-photons) calculated with the Monte Carlo code TRIPOLI-4{sup R} inside different experimental channels (water) located into the OSIRIS periphery and used during the CARMEN-1P irradiation campaign. We will consider Neutrons- Photons-Electrons and Photons-Electrons modes. We will begin by a brief description of the differential-calorimeter device geometry. Then the MCNP5 model used for the calculations of nuclear heating inside the calorimeter elements will be introduced. The energy deposition due to the prompt-gamma, delayed-gamma and neutrons, the neutron-activation of the device will be considered. The different components of the nuclear heating inside the different parts of the calorimeter will be detailed. Moreover, a comparison between KERMA and nuclear energy deposition estimations will be given. Finally, a comparison between this total nuclear heating Calculation and Experiment in graphite sample will be determined. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Removal of Trace Amounts of Co-60 from NPP Waste Water Using UV-C Irradiation, TiO{sub 2} and Co-Treat Inorganic Ion Exchange Media - 16025

The removal of Co-60 from an aqueous nuclear liquid waste obtained from a BWR NPP was tested using a two-step process. In the first phase the liquid waste was oxidized with ultraviolet irradiation at 254 nm (UV-C irradiation) together with a photo-catalyst, TiO{sub 2} P25 AEROXIDE, to degrade the organic complexing agents present in the solution. During some of the experiments it was noticed that TiO{sub 2} P25 AEROXIDE also removed Co-60 from the solution. In the second phase the UV-C irradiated solution was mixed with inorganic ion exchange media, Co-Treat, to remove the rest of Co-60 from the solution. The effect of pH and contact time with UV-C irradiation were evaluated and both had a noticeable effect on the results. Best results were attained when pH of the initial test solution was lowered to 3.5 ± 0.2 before the UV-C irradiation. From the initial Co-60 activity, 2.2 Bq/ml, the activity decreased to 0.03 ± 0.01 Bq/ml after 24 hours of UV-C irradiation together with the photo-catalyst and mixing with Co-Treat. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Development of Surface Grain Relief on Self-Ion Irradiated Copper

The development of surface grain relief has been observed on self-ion irradiated pure polycrystalline copper. However, the mechanistic origin of this feature has yet to be determined. Two possible mechanisms, anisotropic creep relaxation of void swelling induced stresses and crystallographic orientation-dependent sputtering rate differences, are being investigated as the cause of this evolution. This work focuses on characterizing the surface development and investigating the mechanism of orientation-dependent sputtering. Based on the conditions under which these specimens were irradiated, preferential sputtering is not likely to be the cause of the observed grain relief. Surface grain relief can greatly influence the ability to non-destructively detect irradiation-induced effects in polycrystalline metals. Determining the origin of grain relief development is necessary before this surface feature can be used an indicator of irradiation-induced effects in pure copper and other alloys. The development of surface grain relief has been observed on self-ion irradiated pure polycrystalline copper irradiated to 100 dpa. This surface feature was first noted by measuring samples using transient grating spectroscopy and later confirmed using scanning electron microscopy. Calculations of single-crystal sputtering rates show that the difference in sputtering yields between a randomly-oriented solid and a low-index plane is not sufficient to account for the observed height difference. SEM images appear to show a maximum step-height difference of around 100 nm and differential sputtering can only account for differences on the order of 10 nm at most. This determination indicates that anisotropic creep relaxation may be the mechanistic origin of the observed surface features. If this is the case, which will be confirmed by future work, then TG spectroscopy may be an effective tool for the determination of the onset of certain radiation-damage evolution. For example, if the onset of creep relaxation is strongly correlated to a specific level of void swelling and if the surface features present as a result of creep relaxation can be observed using TG spectroscopy, then accurate determinations of particular levels of void swelling may be made. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2019

How do Early Age Geopolymers Respond to Gamma Irradiation Exposure? - 19515

Geopolymers have been proposed by the UK nuclear industry as a potential replacement for the currently accepted blend of Portland cement and supplementary cementitious materials, such as fly ash and blast furnace slag. This preferred grouting system is not ideal for the encapsulation of some problematic waste streams, such as oils and reactive metals. Metakaolin based geopolymers have been selected as the alternative grouting system, as this mitigates the supply risk which is currently associated with the supplementary cementitious materials described above. To allow these systems to be used as an alternative nuclear grouting system, the effect of irradiation on the 'neat' grout needs to be understood. The majority of the literature appears to present studies of the irradiation of aged grouts (3 - 7 days). This study will assess samples that have been irradiated at early age curing (20 hours). The effects of clay type and total dose will also be studied, and a comparison between the effects of rotary and flash calcined clays, irradiated to doses up to 1 MGy will be presented. The thermogravimetry data indicates that irradiation promotes the liberation of water and also the formation of carbonate species. The rotary calcined metakaolin liberates more water than the flash calcined metakaolin. Therefore, it is expected that the flash calcined system is more stable to irradiation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Post-irradiation Experiment on IG-110 and A3-3 Graphites for HTGR - 18494

The first irradiation test of A3-3 graphite matrix and IG-110 graphite were performed in Hanaro during the period of August 2013 through March 2014. The dimensional and weight changes, microstructure, and thermal diffusivity were investigated in different directions of the specimen after the irradiation. Interestingly, the density of the A3-3 matrix was increased by 9.4% due to the weight increase after the irradiation. Moreover, the thermal diffusivity in the radial direction of the A3-3 matrix is almost 3-times higher than that of the axial direction owing to the radially oriented graphite texture shown in the microstructure. Meanwhile, 10-20 {mu}m sized irradiation induced pores are observed in the microstructure of IG-110. The thermal diffusivity measurement of pre- and post-irradiated IG-110 results in a 68% irradiation-dominant decrease of thermal conductivity. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal Diffusivity Measurement of Irradiated SiC-SiC Composite Material

Silicon Carbide fiber and matrix composite (SiC-SiC) is being considered as a fuel cladding candidate because of its high melting point, low susceptibility to corrosion, good thermal and mechanical shock resistance, and low degradation of mechanical properties under irradiation. The thermal properties of SiC-SiC composite depend heavily on the fabrication process, thus cannot be simply referenced from available database. Similarly, the influence of irradiation on SiC-SiC composite thermal properties needs to be measured to provide general trends of property degradation. For ion irradiated samples, the damaged layer is only a few to a few tens of micrometers. It is impossible to directly measure the bulk property change of a SiC-SiC composite because of the thin, damaged layer. This paper focuses on a methodology that is based on microscale measurement of fiber and matrix properties to determine the overall bulk property change due to ion irradiation. Laser-based photothermal reflectance technique is one of the ideal methods to measure local thermal conductivity and diffusivity with high spatial resolution. The main principle of this technique is to heat the solid sample using an intensity-modulated laser, and scan the sample with a constant intensity continuous-wave (CW) laser to detect the phase lag of the thermal wave propagation. On bulk, isotropic material, if the heating laser intensity is modulated as a sine wave, thermal diffusivity of the sample can be derived from the phase lag, scan distance and modulation frequency in a straight forward manner. By using more specific analytical models, thermal conductivity and diffusivity in complicated cases, such as layered structure materials, composite materials, and/or anisotropic materials, are measurable as well. In this work, a new measurement approach of SiC-SiC composite thermal diffusivity is presented. Due to the micrometer spatial resolution of the photothermal reflectance technique, the measurements are performed locally on fibers along both radial and axial directions, and on the matrix. The influence of Silicon ion irradiation (75 MeV, 10 dpa) is then examined by comparing the experimental results of samples before and after irradiation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effects of irradiation damage on the elastic properties and hardness of complex transition metal diborides

Multi-principal component transition metal (TM) diborides represent a class of high-entropy ceramics (HECs) that have received considerable interest in recent years owing to their promising properties for extreme environment applications that include thermal/ environmental barriers, hypersonic vehicles, turbine engines, and next-generation nuclear reactors. While the addition of chemical disorder through the random distribution of TM elements on the cation sublattice has offered opportunities to tailor elastic stiffness and hardness, the effects of irradiationinduced structural damage on the physical properties of these complex materials have remained largely unexplored. To this end, changes in the hardness and elastic moduli of a high-entropy TM diboride (Hf0.2Nb0.2Ta0.2Ti0.2Zr0.2)B2 and three of its quaternary subsets following irradiation with 10 MeV gold (Au) ions to fluences of up to 6 × 10^15 Au cm-2 are investigated at the micrometer and sub-micrometer length-scales via the dispersion of laser-generated surface acoustic waves (SAW) and nanoindentation, respectively. The nanoindentation measurements show that the TM diborides exhibit an initial increase in hardness following irradiation with energetic Au ions, with a subsequent decrease in hardness following further irradiation. One quaternary composition, (Hf1/3Ta1/3Ti1/3)B2, exhibits a notable exception to the trend and continues to exhibit an increase in hardness with ion irradiation fluence. Although differences in the absolute values of the effective elastic moduli obtained from the measured SAW dispersion and nanoindentation are observed (and attributed to microstructural variations at the measurement length-scale), both techniques yield similar trends in the form of an initial reduction and subsequent saturation in the elastic modulus with increasing ion irradiation fluence. The quaternary TM diboride (Hf1/3Ta1/3Ti1/3)B2 again exhibits a departure from this trend. The high-entropy TM diboride (Hf0.2Nb0.2Ta0.2Ti0.2Zr0.2)B2 exhibits the greatest recovery in hardness and modulus when irradiated to high ion fluences following initial changes at low fluence, indicating superior resistance to radiation-induced damage over its quaternary counterparts. Opportunities for designing HECs with superior hardness and modulus for enhanced radiation resistance (compared to their single constituent counterparts) by tailoring chemical disorder and bond character in the lattice are discussed.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

Irradiation Testing of Additively Manufactured Materials for LWR Applications: Alloy 718 and 316L Stainless Steel

The objective of this NSUF Project is to assess the changes in irradiated additively manufactured (AM) material properties as compared to non-irradiated material. Type 316L stainless steel and Alloy 718 samples were produced using Direct Metal Laser Melting (DMLM) fabrication. Materials produced from this fabrication method have several potential applications within the nuclear industry as reactor internal repair parts, fuel debris resistant filters, or fuel spacers within existing light water reactors (LWRs). AM materials have been shown to achieve equivalent mechanical behavior in simulated reactor environments as compared to wrought materials, but have significantly more flexibility when it comes to unique design features. The increased component design flexibility makes these AM materials an attractive choice for both current LWR applications as well as for small modular reactor (SMR) designs. Prior to use of these materials in reactor fleet operation, the industry as a whole must evaluate the effects of irradiation on their material properties. Standard 0.4 inch thick Compact Tension specimens and SSJ3 type tensile bars were neutron irradiated at the Advanced Test Reactor to ~1 dpa for the purpose of performing a variety of mechanical tests in a range of simulated environments applicable to LWRs. For the ductile austenitic Type 316L stainless steel, the irradiated data will be used to confirm that the AM process produces materials with properties that are equivalent to wrought materials under testing conditions applicable to LWR operation. Transmission electron microscopy analysis was also performed in order to understand microstructural and microchemical changes induced in each material in response to neutron irradiation. If possible, data collected from these AM 316L samples will be used to remove fluence limits from specifications of ASME code cases for this alloy, which will give vendors much more flexibility in building future components.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

Mechanical and Microstructural Study of Neutron-Irradiated Novel Alloy Systems

Generation IV nuclear reactors require structural materials capable of withstanding high temperatures, corrosion, and radiation damage. Refractory high entropy alloys (RHEAs) show potential due to their resistance to irradiation damage, reduced void swelling, and microstructural stability. However, most previous studies have focused on thin films and ion irradiation, leaving the neutron irradiation response of bulk RHEAs largely unexplored. Mechanical behavior results show brittle behavior before and after irradiation. The addition of Zr led to an increase in hardness due to Zr-rich precipitates in the bulk microstructure. Lastly, hardness results showed an increase in hardening after irradiation indicating irradiation induced hardening has occurred.

36 - MATERIALS SCIENCE↗
DOE OSTI2017

Effect of tube processing methods on microstructure, mechanical properties and irradiation response of 14YWT nanostructured ferritic alloys

In this research, innovative thermal spray deposition (Process I) and conventional hot extrusion processing (Process II) methods have been used to produce thin walled tubing (~0.5 mm wall thickness) out of 14YWT, a nanostructured ferritic alloy. The effects of processing methods on the microstructure, mechanical properties and irradiation response have been investigated by using scanning electron microscopy (SEM), transmission electron microscopy (TEM) and, micro- and nano-hardness techniques. It has been found that these two processes have a significant effect on the microstructure and mechanical properties of the as-fabricated 14YWT tubes. Even though both processing methods yield the formation of various size Y-Ti-O particles, the conventional hot extrusion method results in a microstructure with smaller, homogenously distributed nano-oxides (NOs, Y-Ti-O particles < 5 nm) with higher density. Therefore, Process II tubes exhibit twice the hardness of Process I tubes. It has also been found that these two tremendously different initial microstructures strongly affect irradiation response in these tubes under extremely high dose ion irradiations up to 1100 peak dpa at 450 °C. The finer, denser and homogenously distributed NOs in the Process II tube result in a reduction in swelling by two orders of magnitude. On the other hand, inhomogeneity of the initial microstructure in the Process I tube leads to large variations in both swelling and irradiation induced hardening. Moreover, hardening mechanisms before and after irradiation were measured and compared with detailed calculations. In conclusion, this study clearly indicates the crucial effect of initial microstructure on radiation response of 14YWT alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2025

MX precipitate behavior in an irradiated advanced Fe-9Cr steel: Helium effects on phase stability

As part of an ongoing series aimed at optimizing Fe-9Cr reduced activation ferritic/martensitic (RAFM) alloys for fusion energy systems, this study explores MX precipitate behavior under dual-ion irradiations, specifically examining correlations between helium transmutation and irradiation-induced damage. Utilizing single and dual-beam ion irradiation, the research explores the combined effects of helium (10–25 appm He/dpa), temperature (400–600 °C), and damage levels (15–100 dpa) on the microstructural evolution of CNA9 steel, a variant of Castable Nanostructured Alloys (CNAs). The study demonstrates that helium co-implantation hinders radiation-enhanced coarsening of MX-TiC precipitates at 500 and 600 °C, maintaining MX-TiC precipitate stability at moderate damage levels (15 dpa) but failing to prevent complete precipitate dissolution at higher damage levels (≥50 dpa) when irradiated at 500 °C. Here, a generalized precipitate stability model suggests that helium-induced suppression of diffusion alters the balance between recoil resolution and back diffusion for MX-TiC precipitates, enhancing the current understanding of precipitate behavior under damage and transmutation simulated dual-ion irradiation conditions.

Characterization↗
DOE OSTI2023

The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels

Embrittlement of light water reactor pressure vessels (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. In this study, the steels were irradiated to a fluence of ~ 1.38 x 10 20 n/cm 2 at ~ 292 °C in the UCSB ATR-2 experiment. In Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promotes precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Amorphization kinetics in strontium titanate at 16 and 300 K under argon ion irradiation

Here, the accumulation of irradiation-induced disorder in SrTiO 3 single crystals irradiated at 16 K with 200 keV Ar ions has been investigated using Rutherford backscattering spectrometry along the < 100 > channeling direction and compared with previous results obtained at 300 K under identical irradiation conditions. As expected, amorphization at 16 K occurs at a much lower fluence than at 300 K due to dynamic recovery of irradiation-induced defects at 300 K. Utilizing a comprehensive damage accumulation model for analysis of the data, irradiation at 16 K results only in the formation of point defects and amorphous pockets, while defect clusters are also formed at 300 K. High defect mobility under irradiation at 300 K tends to promote recombination and clustering of point defects. These results suggest that defect diffusion processes in SrTiO 3 are not thermally active at 16 K.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Shear Punch Testing of Neutron-Irradiated HT-9 and 14YWT

Shear punch mechanical testing was performed on neutron-irradiated HT-9 and 14YWT alloys, comparing the traditional ferritic/martensitic steel with the advanced oxide dispersion strengthened nanostructured ferritic alloy. The samples were irradiated at various temperatures between 380°C and 520°C. In HT-9, on increasing the irradiation temperature, the 1% offset shear stress and maximum shear stress decreased. The advanced 14YWT alloy was less sensitive to irradiation temperature with respect to irradiation hardening. However, it still exhibited irradiation hardening at each temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Neutron irradiation-induced microstructure damage in ultra-high temperature ceramic TiC

Titanium carbide (TiC) is an ultra-high temperature ceramic with potential as a structural material candidate for advanced reactor concepts. However, the irradiation tolerance of TiC is not well understood. Here, we reveal the key irradiation damage microstructure degradation processes in TiC using mixed spectrum neutron irradiations at dose of ~2 displacements per atom (dpa) at temperatures of ~220, 620, and 1115 °C, combined with state-of-art microstructure characterization using transmission electron microscopy (TEM) and high-resolution TEM (HRTEM). The helium injection (~65 atomic parts per million) produced by neutron transmutation also occurred in the sample. TiC was observed to form irradiation-induced interstitial-type dislocation loops and He-stabilized cavities. At 220 and 620 °C, the analysis of the electron diffraction patterns, rel-rod imaging and HRTEM revealed that the dislocation loops were faulted Frank loops with Burgers vector b Frank = 1/3<111> lying on {111} planes. A detailed Burgers vector identification performed by the g.b technique revealed that the dislocation loops forming at 1115 °C were unfaulted, edge-type, with Burgers vectors b = a<100> and a/2<110> with corresponding {100} and {110} habit planes. Using continuum mechanics, we estimated the critical radius at which a faulted dislocation loop transitions to a perfect loop to be 9 nm. Further, no amorphization occurred in TiC under-investigated irradiation conditions while macroscopic swelling under point defect swelling regime was observed. He-stabilized cavities were detected at 1115 °C, but not at lower temperature irradiation. These observations indicate the onset of long-range vacancy migration in TiC occurs between 620 and 1115 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Novel effects of grain size and ion implantation on grain boundary segregation in ion irradiated austenitic steel

Nuclear reactor environments are extreme in nature due to the combination of exposure to corrosive coolant, mechanical stresses, and neutron irradiation damage. This combination of environments can lead to irradiation induced/assisted precipitation, segregation, and embrittlement coupled with enhanced creep and stress corrosion cracking. While nanostructured materials are of interest in a variety of applications due to their unique properties including high mechanical strength, for nuclear applications nanostructured materials have the added benefit of irradiation resistance due to the high volume density of sinks for irradiation-induced defects. In this study, we show unique radiation-induced segregation (RIS) behavior in nanocrystalline 304 stainless steel (SS304). Typical RIS in austenitic steels includes Ni enrichment and Cr depletion within the grain boundaries. In nanostructured SS304 this RIS is reduced. Furthermore, interestingly, some grain boundaries in the nanocrystalline SS304 in this study were found to be enriched in Cr after Fe 2+ ion irradiation at 500 °C up to 50 displacements per atom (dpa). To understand this unique observation, lattice-based atomic kinetic Monte Carlo simulations were performed to investigate the effects of grain size, thermal segregation, as well as injected interstitials (introduced as Fe self-ions). The results indicate that the Cr enrichment at grain boundaries may be caused by combined effects of grain size, self-ion injection and preferential Cr diffusion via the interstitial mechanism. In conclusion, the findings also suggest a possible difference in RIS between ion and neutron irradiations.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Pigment modulation in response to irradiance intensity in the fast-growing alga Picochlorum celeri

Picochlorum celeri has among the fastest photoautotrophic growth rates (~2 h doubling time in optimal conditions) reported to date for a marine alga. This study comprehensively analyzes the levels of Picochlorum celeri photosynthetic pigments, which can reach up to ~13% of the total particulate organic carbon (POC) under light-limiting conditions. The main Picochlorum celeri pigments identified include: chlorophyll a, chlorophyll b, lutein, β-carotene, canthaxanthin, violaxanthin, neoxanthin, zeaxanthin and antheraxanthin. The ketocarotenoid canthaxanthin can accumulate up to 120 mg L -1 in Picochlorum celeri liquid culture; ~12% of it was found in an extracellular polysaccharide matrix. Using a solar-simulating automated photobioreactor, we monitor the photoacclimation of cultures maintained in unshaded conditions (<0.5 μg mL -1 of total chlorophyll) through transitions from high irradiance (1000 μmoles photosynthetically active radiation (PAR) m -2 s -1 ) to low irradiance (60 μmoles PAR m -2 s -1 ), and conversely from low to high irradiance. Canthaxanthin and zeaxanthin accumulation are among the most rapid modulation responses when cultures are shifted from low to high irradiance. The violaxanthin, antherazanthin, and zeaxanthin (VAZ) pool is ~3-fold higher in high-light cultures, suggesting that the VAZ cycle combined with a dramatic reduction in chlorophyll levels are among the major mechanisms used in Picochlorum celeri to efficiently acclimate to high-irradiance levels. Responsive pigment modulation was also observed in denser cultures (~0.65 g L –1 ) grown under a diel cycle (pond-mimicking conditions). Furthermore, this research provides unique insights into the dynamics of pigment modulation and photoacclimation in response to changing irradiance in a biotechnologically promising alga and will inform future pigment engineering strategies to further improve light capture and biomass accumulation.

59 BASIC BIOLOGICAL SCIENCES↗