Post-irradiation Analysis Research Articles

ITER materials irradiation within the D–T neutron environment at JET: post-irradiation radioactivity analysis following the DTE2 experimental campaign

Abstract This work presents the results following the first irradiation of ITER materials samples in a tokamak D--T plasma environment operating at significant fusion power. The materials exposed to this nuclear environment at the Joint European Torus during the DTE2 experimental campaign that took place in 2021 include representative ITER samples from various components such as poloidal field (PF) coil jacket samples, toroidal field coil radial closure plate steels, EUROFER 97 steel, W and CuCrZr materials from the divertor, Inconel-718, CuCrZr, and 316L stainless steel for blanket modules, as well as vacuum vessel forging samples.

The experimental results discussed include high-resolution gamma spectrometry measurements and analysis conducted with the post-irradiated samples, of which there were 68 in total. These samples were exposed through different experimental campaigns, including deuterium, deuterium--tritium and tritium phases. Supporting the analysis were 25 dosimetry foil-based neutron diagnostics and two 'VERDI' neutron spectrometry diagnostics. A further 12 samples for positron annihilation spectroscopy (PALS) were also irradiated. The irradiation of all these samples took place in a long-term irradiation assembly located near the JET vacuum vessel.

The post-irradiation analysis of the ITER material samples has yielded valuable insights into their material activation levels and radiation fields. Comparative assessments between experimental measurements and comprehensive neutronics simulations have demonstrated a significant level of agreement in this work, while also revealing some discrepancies in specific material instances. The data and interpretation from this work not only serve as a robust experimental foundation for enhancing the precision and predictability of neutronics simulation approaches for ITER and next-step devices but also present some opportunities for the refinement of simulation methodologies. In light of these findings, a series of recommendations have been proposed, aimed at improving confidence in nuclear predictions associated with materials that have been exposed to fusion nuclear environments and advancing understanding in this important domain.

Microstructural evolution and hardening of W-ZrC, W-1Re-ZrC and W-3Re-ZrC alloys induced by He ions irradiation

Bulk W-0.5 wt%ZrC (W-ZrC), W-1 wt%Re-0.5 wt%ZrC (W-1Re-ZrC), and W-3 wt%Re-0.5 wt%ZrC (W-3Re-ZrC) alloys were prepared and exposed to He2+ ion irradiation with an incident energy of 500 keV at 400 °C. Post-irradiation analysis revealed the formation of He bubbles, with consistent number density and average size across all alloys. However, the W-3Re-ZrC alloy exhibited a notably higher density of dislocation loops within grain interiors compared to W-ZrC and W-1Re-ZrC. This led to a significant increase in the hardening rate of W-3Re-ZrC to 61.5 %, nearly twice that of the other alloys. By considering the binding energies of He-vacancy, He-He, and Re-He, the formation and growth behavior of irradiation-induced defects, including He bubbles and dislocation loops, were also discussed. This work provides valuable insight into the relationships among Re alloying, He-induced defects, and irradiation hardening in Re-alloyed W materials, thus offering valuable guidance for the design of W-based plasma-facing materials.

Structural evolution of zirconium diboride under gamma irradiation and thermal annealing

Understanding the thermal properties and enhancing the performance of zirconium diboride (ZrB2) under extreme conditions is crucial, especially given its role as a burnable absorber in nuclear fuel pellets and its potential applications in accident-tolerant fuels. This study explores the structural alterations of ZrB2 crystals induced by gamma irradiation and subsequent thermal annealing. ZrB2 crystals were exposed to absorption doses of 300 kGy, 1000 kGy, 1500 kGy, and 3000 kGy using a 60Co gamma source, followed by thermal treatment at 1173 K under vacuum conditions. Also, Monte Carlo simulations were utilized to analyze the energy deposition and distribution of gamma quanta within the material during irradiation, highlighting Compton effects up to 0.23 MeV. X-ray diffraction analysis elucidated the impact of gamma radiation on ZrB2 crystal lattice parameters, space group, and volume expansion coefficient. Post-irradiation analysis revealed a maximum degree of amorphization and investigated mechanisms governing lattice parameter expansion. Thermal annealing at 1173 K significantly enhanced crystallinity, reducing amorphization to 0.2 % at an absorption dose of 3000 kGy and facilitating the restoration of the crystal's columnar structure. This comprehensive investigation provides insights into the intricate interplay between gamma irradiation, thermal processing, and resulting structural changes in ZrB2 nanocrystals, essential for applications in radiation-exposed environments.

Intensity threshold variation method in the post-irradiation analysis of Fluorescent Nuclear Track Detectors for neutron dosimetry

Fluorescent Nuclear Track Detectors (FNTDs) are biocompatible passive detectors and a promising tool for neutron dosimetry applications, such as in-phantom measurements within the scope of ion therapy. Sensitivity differences between individual detectors however complicate the post-irradiation analysis process, making detector comparability and consequently dose assessment challenging. Therefore, improvements in the analysis method are needed to account for the observed sensitivity effects. In this work, the FNTD technology was implemented for neutron dosimetry using a standard Am-241/Be neutron source for irradiation and a confocal laser scanning microscope for detector readout. Acquired images were evaluated in terms of the track number of recoil proton signals behind a polyethylene converter, which was found to be related to dose. This study presents a new methodology for the post-irradiation data analysis, based on the variation of the intensity threshold parameter during the proton track identification process. The procedure was tested within individual detectors in stepwise irradiation experiments, validating the consistency of the proposed methodology. The application of the method to an extended number of FNTDs was able to account for sensitivity impacts and enabled comparability between the different detectors. A calibration curve based on the observed linear relation between dose and the number of identify tracks was obtained, that can be used for future implementation of FNTDs in neutron dosimetry.

Effects of Low-Level Gamma Radiation on Common Nitroaromatic, Nitramine, and Nitrate Ester Explosives.

The aging of high explosives in an ionizing radiation field is not well understood, and little work has been done in the low dose and low dose rate regime. In this study, four explosives were exposed to low-level gamma irradiation from a 137Cs source: PETN, PATO, and PBX 9501 both with and without the Irganox 1010 stabilizer. Post-irradiation analysis included GC–MS of the headspace gas, SEM of the pellets and powder, NMR spectroscopy, DSC analysis, impact sensitivity tests, and ESD sensitivity tests. Overall, no significant change to the materials was seen for the dose and dose rate explored in this study. A small change in the 1H NMR spectrum of PETN was observed and SEM and ESD results suggest a surface energy change in PATO, but these differences are minor and do not appear to have a substantial impact on the handling safety.

Irradiation temperature monitoring with SiC for RPV steel at low fluence

Accurate knowledge of the irradiation temperature is a key concern in irradiations of reactor pressure vessel (RPV) steel. We report results of passive temperature monitoring of RPV steel with SiC. Two un-instrumented capsules containing RPV steel blocks were irradiated in Belgian Reactor 2, at SCK CEN in Belgium. Because un-instrumented capsules were used, the irradiation conditions (gamma heating and irradiation temperature) were calculated. To have experimental verification of the irradiation conditions each capsule contained a passive silicon carbide (SiC) temperature monitor. After irradiation the resistivity and the radiation-induced swelling (lattice parameter) is measured using x-ray diffraction (XRD). These measurements were repeated after annealing treatments to determine the peak irradiation temperature. The best method to determine the peak irradiation temperature with the lowest uncertainty in this study was the resistivity technique. Although lattice parameter measurements could be improved by using material without preferential orientation. Tensile tests of the RPV blocks were compared to a large available database of RPV material with the proper sensitivity towards neutron fluence and irradiation temperature to find the irradiation temperature of the RPV steel. The data showed a discrepancy between the temperature obtained from SiC and the temperature obtained from tensile tests. The temperature differences were discussed and rationalized by finite element modelling with respect to uncertainties in the helium gap between the capsule and the RPV steel block. The aluminium holder accounted for the temperature difference and the use of a steel holder is recommended for future similar irradiations to minimize the temperature difference between SiC and tensile specimens . In-depth analysis showed that when the irradiation temperature dropped at the end of the irradiation for a considerable time, the SiC temperature monitor had a memory effect. In this case the post-irradiation resistivity analysis showed two peak irradiation temperatures with a region of constant resistivity between the two.

Genetic Responses of Metabolically Active Limnospira indica Strain PCC 8005 Exposed to γ-Radiation during Its Lifecycle.

Two morphotypes of the cyanobacterial Limnospira indica (formerly Arthrospira sp.) strain PCC 8005, denoted as P2 (straight trichomes) and P6 (helical trichomes), were subjected to chronic gamma radiation from spent nuclear fuel (SNF) rods at a dose rate of ca. 80 Gy·h−1 for one mass doubling period (approximately 3 days) under continuous light with photoautotrophic metabolism fully active. Samples were taken for post-irradiation growth recovery and RNA-Seq transcriptional analysis at time intervals of 15, 40, and 71.5 h corresponding to cumulative doses of ca. 1450, 3200, and 5700 Gy, respectively. Both morphotypes, which were previously reported by us to display different antioxidant capacities and differ at the genomic level in 168 SNPs, 48 indels and 4 large insertions, recovered equally well from 1450 and 3200 Gy. However, while the P2 straight type recovered from 5700 Gy by regaining normal growth within 6 days, the P6 helical type took about 13 days to recover from this dose, indicating differences in their radiation tolerance and response. To investigate these differences, P2 and P6 cells exposed to the intermediate dose of gamma radiation (3200 Gy) were analyzed for differential gene expression by RNA-Seq analysis. Prior to batch normalization, a total of 1553 genes (887 and 666 of P2 and P6, respectively, with 352 genes in common) were selected based on a two-fold change in expression and a false discovery rate FDR smaller or equal to 0.05. About 85% of these 1553 genes encoded products of yet unknown function. Of the 229 remaining genes, 171 had a defined function while 58 genes were transcribed into non-coding RNA including 21 tRNAs (all downregulated). Batch normalization resulted in 660 differentially expressed genes with 98 having a function and 32 encoding RNA. From PCC 8005-P2 and PCC 8005-P6 expression patterns, it emerges that although the cellular routes used by the two substrains to cope with ionizing radiation do overlap to a large extent, both strains displayed a distinct preference of priorities.

Post-Irradiation Analysis of Additively Manufactured Stainless Steel 316L Specimens

Post-Irradiation Analysis of Additively Manufactured Stainless Steel 316L Specimens

Mapping the Future of Particle Radiobiology in Europe: The INSPIRE Project

Particle therapy is a growing cancer treatment modality worldwide. However, there still remains a number of unanswered questions considering differences in the biological response between particles and photons. These questions, and probing of biological mechanisms in general, necessitate experimental investigation. The Infrastructure in Proton International Research (INSPIRE) project was created to provide an infrastructure for European research, unify research efforts on the topic of proton and ion therapy across Europe, and to facilitate the sharing of information and resources. This work highlights the radiobiological capabilities of the INSPIRE partners, providing details of physics (available particle types and energies), biology (sample preparation and post-irradiation analysis), and researcher access (the process of applying for beam time). The collection of information reported here is designed to provide researchers both in Europe and worldwide with the tools required to select the optimal center for their research needs. We also highlight areas of redundancy in capabilities and suggest areas for future investment.

Monitoring the temperature of reactor experiments using radiation-induced swelling in SiC

Temperature monitoring of reactor experiments is an important quality requirement. High-purity silicon carbide (SiC) can be used to provide such data through post-irradiation material analysis. This method stems from radiation defect concentrations’ dependence on irradiation temperature. Irradiation temperature can be determined by measuring irradiation-induced property changes, for example lattice spacing distribution, dimensions, electrical resistivity, etc.—after isochronal annealing. Such methods are time-consuming, since multiple steps must be performed in a serial manner. We investigated the possibility of using an alternative method that does not require multiple heat treatments. Instead, peak irradiation temperature is determined by a calibration curve comparing pre- and post-irradiation sample dimensions. The experimental results demonstrate the feasibility of this approach in the irradiation conditions described below. This approach can also be used to measuring changes in other properties mentioned above.

Radiation damage from energetic particles at GRad-level of SiO[formula omitted] fibers of the Large Hadron Collider ATLAS Zero-Degree Calorimeter (ZDC)

Core SiO2 quartz fibers of the Large Hadron Collider (LHC) ATLAS Zero-degree Calorimeter (ZDC) are expected to experience integrated doses of a few giga-Rad (Grad) at their closest position to the LHC beam. An array of fibers was irradiated with 200 MeV protons and spallation-generated mixed spectra (primarily fast neutrons) at the Brookhaven National Laboratory (BNL) Linac. Specifically, 1 mm- and 2 mm-diameter quartz (GE 124) rods of 50 mm length were exposed to direct 200 MeV protons leading to peak integrated dose of ∼28 Grad (∼0.28 GGy). Exposure of 1 mm-diameter SiO2 fibers to a neutron flux was also achieved in the spallation field generated by 128 MeV protons. In the post-irradiation analysis, the quartz fiber transmittance was evaluated as a function of the absorbed dose. Significant degradation of the transmittance and increased radiation damage of the material were observed. Microscopic evaluation of the fibers revealed extensive micro-structural damage and irradiation-induced defects. The measurements revealed that a threshold fluence (∼2.6 1016 p/cm2) or dose of ∼10 Grad (0.1 GGy) appears to exist beyond which light transmittance drops below 10%. Also observed is that fiber transmittance loss increased drastically with SiO2 fiber diameter (1 mm vs. 2 mm diameter). This is attributed, in part, to the earlier lateral leakage from the 1 mm fiber of knock-on electrons and primary protons implying that more damage-inducing protons travel within the bulk of the 50 mm long 2-mm fibers. While Monte Carlo simulations performed tend to support such assumption, future experiments and sensitivity studies are envisioned to address the fiber diameter influence on degradation.

Current Transport Mechanism for Heavy-Ion Degraded SiC MOSFETs

High sensitivity of silicon-carbide (SiC) power MOSFETs has been observed under heavy-ion irradiation, leading to permanent increase of drain and gate leakage currents. The electrical postirradiation analysis confirmed the degradation of the gate oxide and the blocking capability of the devices. At low drain bias, the leakage path is formed between drain and gate, while at higher bias the heavy-ion-induced leakage path is mostly from drain to source. An electrical model is proposed to explain the current transport mechanism for heavy-ion degraded SiC power MOSFETs.

Defect evolution in Ni and NiCoCr by in situ 2.8 MeV Au irradiation

The evolution of radiation-induced defects in Ni and the single-phase concentrated solid solution alloy, NiCoCr, were investigated during in situ 2.8 MeV Au ion irradiation and post-irradiation analysis using transmission electron microscopy. Compared to Ni, both the size and area density of defect clusters decreased in NiCoCr under the same irradiation conditions, suggesting that the chemical complexity, i.e., randomness of lattice site occupations, of NiCoCr suppressed radiation-induced damage. One-dimensional glide of defect clusters was observed in Ni but not in the NiCoCr alloy. The structural nature of small defect clusters in NiCoCr were further investigated using high-angle annular dark field scanning transmission electron microscopy.

Determination of 1-keV to 1-MeV Neutron Flux by Radiative Capture (n, $\gamma $ ) on92,94 Zr Dosimeters

The purpose of this paper is to demonstrate 92,94Zr application in neutron reactor dosimetry. We aim to detect neutron flux in 1-keV to 1-MeV energy range by irradiating 92,94Zr samples under boron nitride (BN) filter. The presence of BN is needed in order to cut off thermal neutron contribution. This paper describes Zr irradiation and experimental results that were obtained by $\gamma $ spectrometry and accelerator mass spectrometry. In parallel, Zr irradiation is modeled with Monte Carlo-based code for neutron transport—TRIPOLI-4v10 developed by CEA (Saclay). Experimental values of 95Zr activity and 93Zr/92Zr ratio are compared to calculation results. Since crosssection values are necessary for the neutron flux determination, we present here the comparison between several nuclear data libraries for radiative capture on 92,94Zr. Uncertainties for calculation/experience (C/E) ratios are obtained combining both experimental and nuclear data ones. This paper shows the feasibility of 1-keV to 1-MeV neutron detection by using 92,94Zr (n, $\gamma $ ) reaction and introduces a postirradiation analysis for reactor dosimetry based on accelerator mass spectrometry. It also reveals that more precise evaluation of 92,94Zr (n, $\gamma $ ) cross section is required in order to improve C/E values.

Proton irradiation effects on beryllium – A macroscopic assessment

Beryllium, due to its excellent neutron multiplication and moderation properties, in conjunction with its good thermal properties, is under consideration for use as plasma facing material in fusion reactors and as a very effective neutron reflector in fission reactors. While it is characterized by unique combination of structural, chemical, atomic number, and neutron absorption cross section it suffers, however, from irradiation generated transmutation gases such as helium and tritium which exhibit low solubility leading to supersaturation of the Be matrix and tend to precipitate into bubbles that coalesce and induce swelling and embrittlement thus degrading the metal and limiting its lifetime. Utilization of beryllium as a pion production low-Z target in high power proton accelerators has been sought both for its low Z and good thermal properties in an effort to mitigate thermos-mechanical shock that is expected to be induced under the multi-MW power demand.To assess irradiation-induced changes in the thermal and mechanical properties of Beryllium, a study focusing on proton irradiation damage effects has been undertaken using 200 MeV protons from the Brookhaven National Laboratory Linac and followed by a multi-faceted post-irradiation analysis that included the thermal and volumetric stability of irradiated beryllium, the stress-strain behavior and its ductility loss as a function of proton fluence and the effects of proton irradiation on the microstructure using synchrotron X-ray diffraction. The mimicking of high temperature irradiation of Beryllium via high temperature annealing schemes has been conducted as part of the post-irradiation study. This paper focuses on the thermal stability and mechanical property changes of the proton irradiated beryllium and presents results of the macroscopic property changes of Beryllium deduced from thermal and mechanical tests.

In-situ TEM studies of 150 keV W+ ion irradiated W and W-alloys: Damage production and microstructural evolution

In-situ irradiations with 150 keV W+ ions have been performed on W and W-5wt.% (Re; Ta; V) alloys in a comprehensive study of the influences of irradiation temperature Tirr, dose, alloying elements and grain orientations on radiation damage production and microstructural evolution. For Tirr between 30 K and 1073 K, the first observable defects in pure W appeared at doses ≤0.01 dpa, and were most likely vacancy loops, with Burgers vectors predominantly of type b = ½ <111>. With increasing Tirr, the retained defect concentration decreased strongly and the maximum cluster size increased from ∼1300 point defects at 30 K to ∼2300 point defects at 1073 K. At all irradiation temperatures, the evolution of damage microstructures with dose from 0.1 to 1.0 dpa involved defect cluster migration, with mutual elastic interactions often leading to spatial inhomogeneities and loop reactions. In pure W, spatial ordering of loops was observed at doses >0.4 dpa and Tirr≥773 K in grains close to z = <001>. No such ordering was found in similar grain orientations for the W-(Re; Ta) alloys, but it was found in the non-z = <001> grains. Post-irradiation analysis on W and W-5 wt% (Re; Ta) at 1.0 dpa showed that ½ <111> and <100> loops of both vacancy and interstitial type were present, at number densities ∼ 1015 loops m−2. In all cases ½ <111> loops were dominant, the fraction of these with interstitial nature increased with Tirr, and the proportion of <100> loops decreased with increasing Tirr. Compared with pure W, microstructures in the W-(Re; Ta) alloys exhibited higher loop number densities and evolved more quickly with increasing dose towards damage saturation.

Post-irradiation analysis of an ISOLDE lead-bismuth target: Stable and long-lived noble gas nuclides

We measured the isotopic concentrations of long-lived and stable He, Ne, Ar, Kr, and Xe isotopes in a sample from a lead-bismuth eutectic target irradiated with 1.0 and 1.4 GeV protons. Our data indicate for most noble gases nearly complete release with retention fractions in the range of percent or less. Higher retention fractions result from the decay of long-lived radioactive progenitors from groups 1, 2, or 7 of the periodic table. From the data we can calculate a retention fraction for 3H of 2–3%. For alkaline metals we find retention fractions of about 10%, 30%, and 50% for Na, Rb, and Cs, respectively. For the alkaline earth metal Ba we found complete retention. Finally, the measured Kr and Xe concentrations indicate that there was some release of the halogens Br and I during and/or after the irradiation.

Determination of Neutron Spectra Within the Energy of 1 keV to 1 MeV by Means of Reactor Dosimetry

The standard procedure for neutron reactor dosimetry is based on neutron irradiation of a target and its post-irradiation analysis by Gamma and/or X-ray spectrometry. Nowadays, the neutron spectra can be easily characterized for thermal and fast energies (respectively 0.025 eV and $ > 1~\hbox{MeV}$ ). In this paper we propose a new target and an innovating post-irradiation technique of analysis in order to detect the neutron spectra within the energy of 1 keV to 1 MeV. This paper will present the calculations performed for the selection of a suitable nuclear reaction and isotope, the results predicted by simulations, the irradiation campaign that is proposed, and the post-irradiation technique of analysis.

Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today׳s precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

Measurement of Volatile Radionuclides Production and Release Yields followed by a Post-Irradiation Analysis of a Pb/Bi Filled Ta Target at ISOLDE

A crucial requirement in the development of liquid-metal spallation neutron target is knowledge of the composition and amount of volatile radionuclides that are released from the target during operation. It is also important to know the total amount produced, which could be released if there was an accident. One type is the lead-bismuth eutectic (LBE) target where different radionuclides can be produced following interaction with a high-energy proton beam, notably noble gases (Ar, Kr, Xe isotopes) and other relative volatile isotopes such as Hg and At. The results of an irradiation experiment performed at ISOLDE on a LBE target are compared with predictions from the MCNPX code using the latest developments on the Liège Intranuclear Cascade model (INCL4.6) and the CEM03 model. The calculations are able to reproduce the mass distribution of the radioisotopes produced, including the At production, where there is a significant contribution from secondary reactions. Subsequently, a post-irradiation examination of the irradiated target was performed. Investigations of both the tantalum target structure, in particular the beam window, and the lead-bismuth eutectic were performed using several experimental techniques. No sign of severe irradiation damage, previously observed in other ISOLDE targets, was found.