Neutron Irradiation Damage Research Articles

Silicon ion radiation as a viable surrogate for emulating neutron radiation damage in silicates

Nuclear power plants are aging around the world, and a precise assessment of irradiation damage in their components is needed. One key component, concrete, and specifically the silicates in its aggregates, can undergo significant expansion upon neutron radiation, which can lead to cracking and, ultimately, structural failure. However, assessing and predicting the extent of damage via neutron radiation is challenging due to reasons such as residual radioactivity and, most importantly, the high time involved. Here, we evaluate whether ion radiation can be a viable surrogate. Specifically, by employing Si2+ ion radiations and a comprehensive multi-modal imaging protocol, we report mineral-specific responses for key silicates such as quartz, albite, anorthite, and microcline. We find that 10 MeV Si2+ ions result in mineral expansions that are remarkably comparable to neutron radiation equivalent expansions (R2 = 0.86, RMSE = 1.29%), opening up pathways towards rapid assessment of silicates subject to irradiation.

Correlative Characterisation of Neutron-Irradiation Damage Induced in Zr Alloys

Correlative Characterisation of Neutron-Irradiation Damage Induced in Zr Alloys

Positron annihilation spectroscopy evaluation of using proton irradiation as a surrogate for early neutron radiation damage

A positron annihilation spectroscopy investigation of early stages of radiation-induced damage in Chinese f/m (ferritic/martensitic) steel SIMP and strengthened variant of f/m steel ODS (oxide dispersion strengthened) EUROFER was conducted. Studied materials were irradiated by neutrons at the LVR-15 research reactor to 0.12 dpa at an average temperature of 260 ± 20 °C, and implanted by protons at the 6 MV Tandetron Accelerator of the Slovak University of Technology, to 0.0196 dpa at room temperature (RT) and 300 °C. The potential of using proton irradiation as a surrogate for early neutron radiation damage was investigated. A lifetime spectrometer and Doppler spectrometer was used to obtain simultaneously positron lifetime (PALS) and Doppler broadening spectra (DBS), which provided experimental data used to characterize the vacancy-type defect and their interaction with H atoms in the studied materials. The obtained results show an increase of mean positron lifetime (MLT) in the proton-implanted samples at RT when compared with high temperature irradiation. Similar trend was observed when comparing vacancy concentration of different irradiation conditions. The impact of temperature on mobility and recombination of defects, as well as effect of vacancy stabilization by H atoms, has been addressed. The effect of ODS microstructure on results has been examined. A methodology of comparing various irradiation conditions has been proposed, where the vacancy concentrations have been normalized by the dpa values. The potential application of proton implantation for evaluation of early stages of neutron damage was found to be promising.

Evaluation of neutron radiation damage in the VVER-1200 reactor pressure vessel

In nuclear reactors, the structural materials of reactor pressure vessel (RPV) are damaged by the radiation that produced by fission reactions. Neutron irradiation damage is one of the most critical factors for the degradation of the reactor pressure vessel (RPV). A displacement per atom (dpa) is a standard measurement for calculating neutron-and gamma-induced radiation damage in materials. The aim of this paper was to evaluate the neutrons radiation damage in the reactor pressure vessel of VVER-1200. In this paper, neutron damage factors, including fast neutron fluxes above 1 MeV, above 0.5 MeV, above 0.1 MeV, the damage rate (dpa/s), and Integrated damage [DPA for one effective full power year (EFPY)] were investigated for VVER-1200/AES-2006 reactor pressure vessel (RPV). The calculations are performed by using NJOY code together with the MCNP6 code at beginning of Cycle (BOC). The cross section library used in present calculations is based on ENDF/B-VII.1 library (ENDF71x). The NJOY code was used for generating DPA cross sections in an ACE formatted file that suitable for using in the MCNP6 code. The MCNP6 code was used for simulating the VVER-1200 reactor core and to calculate neutron DPA rates. MCNP6 simulation of the VVER-1200 reactor core identified the area where the RPV neutron radiation is maximized. The integrated damage for a full effective year was evaluated to be 2.128E-04 DPA/EFPD for the VVER-1200 RPV.

Development of potassium doped tungsten plate for fusion reactor applications

Since fusion reactors including DEMO and beyond are expected to be operated for a longer period of time than ITER, plasma-facing materials (PFM) will suffer by longer term heat load and non-negligible neutron irradiation damage. Therefore, for tungsten (W) materials as PFM, improvement of low temperature brittleness and suppression of recrystallization embrittlement and neutron irradiation embrittlement are required. To address these issues, various modified W materials have been developed by applying alloying with solid solution elements and second-phase dispersion techniques such as solid particles and bubbles. As a result, potassium-doped W (K-doped W), in which K bubbles are dispersed, was found to be effective for all of the above issues. However, the materials used in these evaluations were relatively small and not large enough to fabricate, for example, ITER divertor monoblocks (12 mm thick). In the present study, K-doped W plate with a sufficient thickness (more than 12 mm) for application to fusion reactor components like divertor monoblocks were developed and the fundamental properties were evaluated.

Deuterium reclamation from C-Si codeposits using thermo-oxidation

SiC exhibits a remarkable resistance to neutron irradiation damage and, being a low-Z material, is seen as a potential material candidate for plasma-facing components in magnetic confinement fusion reactors. The current work investigates the reclamation of deuterium from C-Si codeposits produced by sputter-deposition at temperatures from 300 K to 700 K using thermo-oxidation at 350 °C and 400 °C (623 or 673 K). While initial D content was found to be close to that of similarly-produced pure carbon codeposits, hydrogen removal from the investigated C-Si codeposits was significantly reduced comparatively. This reduced removal behaviour was found to follow closely with that observed for B-C codeposits from the DIII-D tokamak. This suggests that hydrogen removal via thermo-oxidation, where a non-carbon inclusion is known to form a stable oxide, is unlikely to be effective.

Microstructure refinement and enhanced mechanical properties of wire arc additively manufactured RAFM steel via post heat treatments

Reduced activation ferritic/martensitic (RAFM) steel with high strength and neutron radiation damage resistance is deemed one of the promising structural materials for fusion blankets. In this study, RAFM steel tube was prepared through wire arc additive manufacturing, and the evolution of phase composition, microstructural features and mechanical properties during the as-built and heat treatment processes were investigated. The results show that the as-built state is predominantly composed of α-Fe, and the microstructure's morphology from bottom to top consists of alternating columnar grain regions, equiaxed coarse grain regions, and fine grain regions. Normalizing and tempering treatment eliminates the heterogeneous microstructure, resulting in the presence of M23C6 carbides enriched with Cr and W, as well as MX phases enriched with Ta and V. Partitioning heat treatment results in the formation of ultrafine martensitic, quenched martensite and bainitic microstructures. The mechanical properties of wire arc additively manufactured RAFM steel are influenced by variations in microstructure. The coarse-grained region formed in the as-built state represent weak links. Normalizing and tempering treatment reduces the strength while improving ductility. Partitioning heat-treated RAFM steel exhibited a high strength of 1.1 GPa and an elongation of approximately 10.9%. This study provides a method of combining additive manufacturing with heat treatment to enhance the strength and ductility of RAFM steel simultaneously.

Neutron Radiation Damage on a Planar Segmented Germanium Detector

Neutron Radiation Damage on a Planar Segmented Germanium Detector

TEM study of neutron radiation damage in tungsten

TEM study of neutron radiation damage in tungsten

Effect of low dose irradiation of heavy ion on electrochemical corrosion and IASCC behavior of austenitic steel

The impacts of low dose irradiation on the behavior of electrochemical corrosion and irradiation assisted stress corrosion cracking for 321 stainless steel were studied using Fe2+ ion irradiation to simulate neutron radiation damage in primary circuit environment of pressurized water reactor. Low dose irradiation can improve the pitting resistance and reduce the cracking tendency of the alloy in B-Li solution to a certain extent, which was related to the δ phase content on the near-surface of the sample: The higher δ phase content on the near-surface of the 2 dpa irradiated sample was observed by grazing incident X-ray diffraction. In addition, the pits was significantly increased near micro-cracks for the unirradiated sample, indicating that the existence of pits induced the initiation of cracks. The research results provided an important reference for the failure mechanism of irradiation assisted stress corrosion cracking of core components.

Assessment of neutron irradiation effects via PKA spectra, displacement damage, and gas production: Application to reactor pressure vessel

Neutron irradiation damage is a major challenge for materials employed in radiation environments, such as Reactor Pressure Vessels (RPV). Based on a determined neutron flux, the present study assesses the neutron irradiation effects via Primary Knock-on Atoms (PKAs) production, the consequent displacement damage, and gas production. The complete PKA spectra including all minor isotopes from nuclear reactions are calculated and used to estimate the displacement damage and gas production. At the inner surface of the RPV, which suffers the strongest neutron irradiation, it is found that some minor PKAs have smaller atomic numbers and higher kinetic energies than dominant isotopes in RPV. These PKAs have much deeper ranges in RPV steel and may thus produce important damage other than point defects. The displacement damage is calculated using SRIM simulations and the corresponding corrections to ensure consistency with the NRT-dpa concept. Notably, the displacement damage in the RPV should be ∼23% higher than the prediction by the NRT-dpa formula, which cannot be directly applied to PKAs different from the target atom. The 4.1 H-appm/dpa and 0.2 He-appm/dpa gas productions at the RPV inner surface are higher than that in the innermost 1/4 thickness by a factor of 2 and 1.3, respectively.

Comparing neutron and helium ion irradiation damage of REBa2Cu3O coated conductor using x-ray absorption spectroscopy

Understanding the effects of high energy neutron damage on REBa2Cu3O (REBCO) coated conductor is of vital importance for the design of the magnetic confinement systems for compact nuclear fusion power plants. However, neutron irradiation campaigns can only be carried out in a few facilities, and the experiments are very slow and expensive partly because the samples become radioactive. Ion irradiation provides an easily accessible alternative route to studying the effects of radiation on high temperature superconductors, which not only increases the volume of technical data that can be obtained but also enables more complex experiments such as in situ cryogenic irradiation. The question is, does ion damage offer a good proxy for neutrons? Here we use high energy resolution fluorescence detected x-ray absorption spectroscopy to probe the effects of fast neutron irradiation on the local environment around the copper ions in the REBCO layer of coated conductor tapes. We find that the spectral changes are similar to those induced by helium ion irradiation, suggesting that both projectiles produce the same types of structural defect in the REBCO lattice, although there is some evidence of an additional type of defect present in the sample heavily damaged by He+ ion irradiation. It is also shown that the linear degradation of superconducting transition temperature (T c ) of coated conductors with the calculated number of displacements per atom occurs at the same rate for neutrons and helium ions. Together these results provide new evidence suggesting that helium ions can emulate neutron point defect damage in REBCO high temperature superconductor reasonably well, increasing confidence that helium ions could be used as a useful proxy for neutrons in future experiments.

The mechanism of the irradiation synergistic effect of silicon bipolar junction transistors explained by multiscale simulations of Monte Carlo and excited-state first-principle calculations.

Neutron and γ-ray irradiation damages to transistors are found to be non-additive, and this is denoted as the irradiation synergistic effect (ISE). Its mechanism is not well-understood. The recent defect-based model [Song and Wei, ACS Appl. Electron. Mater. 2, 3783 (2020)] for silicon bipolar junction transistors (BJTs) achieves quantitative agreement with experiments, but its assumptions on the defect reactions are unverified. Going beyond the model requires directly representing the effect of γ-ray irradiation in first-principles calculations, which was not feasible previously. In this work, we examine the defect-based model of the ISE by developing a multiscale method for the simulation of the γ-ray irradiation, where the γ-ray-induced electronic excitations are treated explicitly in excited-state first-principles calculations. We find the calculations agree with experiments, and the effect of the γ-ray-induced excitation is significantly different from the effects of defect charge state and temperature. We propose a diffusion-based qualitative explanation of the mechanism of positive/negative ISE in NPN/PNP BJTs in the end.

Enhanced backscatter and unsaturated blue wavelength shifts in F-doped fused silica optical fibers exposed to extreme neutron radiation damage

Amorphous fused silica (a-SiO2) optical fibers, with and without inscribed Bragg gratings, were interrogated using infrared (∼1550 nm) optical backscatter reflectometry during ∼25 days of intense neutron irradiation to a fast (energy > 0.1 MeV) neutron fluence of ∼1021 n/cm2, or ∼1.5 atomic displacements per atom. The reflected light amplitudes in Ge-doped core telecommunications fiber dropped below detection limits (>15 dB attenuation) within 3 days of irradiation (∼1020 fast n/cm2). Amplitudes from a pure silica core, F-doped silica cladding fiber reached equilibrium levels ∼1.5–2 dB higher than pre-irradiation values, whereas Bragg gratings inscribed in the same fiber using a femtosecond laser (point-by-point method) suffered > 45 dB attenuation. Blue wavelength shifts were initially consistent with previous radiation-induced compaction models but increased linearly with increasing neutron fluence (no evidence of saturation) beyond 1020 n/cm2 and exceeded 0.6%, which corresponds to >1,000 °C drift if used to measure temperature changes.

Humidity effects on neutron irradiated beryllium

This study focuses on neutron-irradiated beryllium interactions with air of differing humidity, simulating water cooled fusion reactor malfunction.Beryllium in the form of ∼ 1 mm ∅ pebbles has been used in the study. Both non-irradiated and neutron-irradiated pebbles were thermally treated and studied using thermogravimetric/differential thermal analysis in dry airflow with relative humidity (RH) < 5 % and humid > 90% RH.Surface structure and composition was investigated with scanning electron microscopy (SEM) and energy dispersion X-ray spectrometry (EDX). Tritium release was monitored using a tritium monitor.Results show that elevated RH coupled with neutron irradiation damage on beryllium yields the largest mass increase of 180% (in relation to the initial mass) compared to non-irradiated Be treated in RH < 5% which resulted in 3 % mass increase. SEM and EDX revealed a substantial number of cracks as well as the formation of irregular shape micro-/nanoparticles of BeO. Tritium release took place at temperatures above ∼ 1040 K, reaching a peak after 1263 K. Overall tritium inventory was estimated to be up to 1.08∙1017 T atoms/g.

Investigation of the radioprotective properties of Zingerone, Ibuprofen, and Sulfasalazine against neutron radiation damage

Nuclear radiation (neutron, gamma X rays, etc.) is widely used in power generation, industry, medicine, research, and agriculture. Exposure to ionizing or indirect ionizing radiation can cause mutagenetic cell death depending on the dose and duration. It is important to reduce the carcinogenesis of cancer patients undergoing radiotherapy treatment and people working in radiation fields. Neutrons have wide applications in neutron activation analysis (NAA), neutron capture therapy (NCT), neutron radiography (NR), neutron capture, neutron conversion doping (NTD) sample irradiation, and neutron diffraction (ND). In these applications, protective substances are needed to stop the neutrons from entering the body and to reduce their effects. In this work, the use of zingerone, ibuprofen, and sulfasalazine compounds as active ingredients in a neutron protective drug was investigated. Some important neutron attenuation parameters such as an effective removal cross section, mean free path, half-value layer, and transmission number were theoretically calculated with the Monte Carlo simulation GEANT4 code for fast and epithermal neutrons.The 241Am-Be fast neutron source and a BF3 gaseous detector were used to determine the neutron absorption ability of all of the samples. All of the results were compared with paraffin, which has a good shielding capacity. According to the results, it was found that every examined sample had a fast and epithermal neutron absorption capacity. It reported that the dose amount from the source was absorbed by the samples at rates of 33.37% (DAI1), 34.85% (DAI2), and 34.21% (DAI3). It was determined that the DAI2 sample in particular had a high absorption ability ratio compared to other samples. Because this sample have high hydrogen content than others samples. So, it was shown that these compounds can be used as active substances to make a radioprotective drug.

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. This study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH2, Mg(BH4)2, and ZrH2 could reduce the total heating on the magnet and also reduce the TBR.

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 Fe2+ 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. The findings also suggest a possible difference in RIS between ion and neutron irradiations.

Resonance self-shielded multigroup atomic displacement cross section for assessing neutron irradiation damage

Accurate calculation of neutron-induced displacement damage is a key step for the subsequent investigations on neutron irradiation damage. Due to the use of multigroup displacement damage cross sections, the resonance self-shielding effect of neutron-induced nuclear reaction must be taken into account. In addition to the method proposed by the authors in 2019, two additional methods are proposed in the present work to consider the self-shielding effect on the calculation of neutron irradiation-induced displacement damage. The two newly proposed methods allow directly using self-shielded multigroup nuclear reaction cross sections without infinite-dilution cross sections, even though they are not as consistent as the first one in the case of weak resonance self-shielding effect. The resulting resonance self-shielding corrections on the displacement damage of 56Fe in a typical sodium-cooled fast reactor are quite close for the three different methods. The slight differences (within ∼ 2 %) are much smaller than the self-shielding corrections, which are about −11.3 % for the dominant neutron elastic scattering-induced damage and −10.4 % for the total displacement damage.

Neutron capture of UO2 targets prepared by spin-coating assisted combustion synthesis

Two uranium dioxide (UO2) targets of (414 ± 23) nm and (1092 ± 93) nm thicknesses were prepared on 6061 aluminum alloy and puratronic grade aluminum backing materials. The targets were deposited with a novel method combining spin coating and solution combustion synthesis (SCS). The target layers consisted of small (3–7 nm) UO2 grains and uniformly distributed ultra-small (1–3 nm) pores. The prepared targets were tested at the Los Alamos National Laboratory’s LANSCE facility for neutron irradiation damage and suitability for neutron capture experiments. The samples showed no signs of target material loss after the irradiation. However, irradiation caused a significant increase in the grain size (4–10 nm), as well as upward mass diffusion and coalescence of the pores due to the thermal spikes. The magnesium in the aluminum 6061 alloy backing also diffused into the UO2 layer during neutron irradiation. The structural changes in the target after the irradiation do not affect the data from neutron capture. The new method can be used more broadly to prepare other actinide targets for nuclear physics experiments.