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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Fission gas trapped in Chornobyl fuel microparticles reveals details of reactor operations

The isotopic ratios of fission gas would provide important source information of a nuclear fuel sample found in the environment. However, it is believed that during a reactor accident like Chornobyl all fission gas is lost and that the radioactive particles found in the Chornobyl Exclusion Zone today are depleted in gases by the initial explosion and subsequent fire. We disprove this hypothesis by detection and analysis of trapped krypton and xenon in these particles. Our analysis of krypton and xenon isotopes by noble gas mass spectroscopy in combination with resonance ionization mass spectrometry establishes that important information about reactor operations like age, neutron flux and plutonium fission fraction can still be reconstructed from individual micrometer-sized particles even after decades of weathering in the environment.

Chornobyl↗
DOE OSTI2021

Radiation Effects Related to Repaired BWR Core Shrouds

The material properties of the core shroud and supporting tie rod components are important factors for Boiling Water Reactor (BWR) life extension considerations. Given the variability of nickel and boron content found in 304 stainless steel shroud plates, and XM-19 and Inconel X-750 tie rod materials, this paper combines the latest spectral information in the vicinity of the core shroud to calculate hydrogen and helium gas production limits as well as atomic displacement damage spanning 80 full-power years (FPY) using the SPECTER code. Results indicate that, whereas boron has little effect on the atomic displacement damage, helium production from boron can be important and must be included in weldability assessments. Since the boron concentration is uncertain, helium generation is calculated assuming a composition range based on specifications and reported measurements. The results show that helium production caused by transmutations in the 304 SS shroud material are dominated by the concentration of boron and will have a significant impact on weldability after only a few years of service. For tie rods, the susceptibility to irradiation-assisted stress corrosion cracking (IASCC) has been assessed by determining helium production in Inconel X-750 IASCC specimens irradiated in the Advanced Test Reactor (ATR) reactor and correlating with the helium produced in a BWR. As a tie rod material in a high thermal neutron flux environment, Inconel X-750 is more likely to be IASCC sensitive during service in a BWR due to helium than the XM-19 stainless steel alloy because of a greater nickel content and boron concentration. Weldability of the core shroud plate material, IASCC of the tie rod materials, and the gamma photon contribution to radiation damage are discussed.

radiation damage in materials, BWR↗
DOE OSTI2021

Embedded Sensors in Additively Manufactured Silicon Carbide

Silicon carbide (SiC) components are being considered for a wide range of nuclear applications due to their high-temperature strength retention, low neutron absorption, chemical inertness, and dimensional stability under neutron irradiation. However, machining and joining of SiC components have traditionally limited its application to relatively simple geometries. Recent work has demonstrated additive manufacturing of complex, high-purity, crystalline SiC components using a combination of binder jet printing and densification via chemical vapor infiltration (CVI). The process lends itself to embedding of fuel, absorbers, moderators, and sensors at strategic locations within a component. The latter could allow for enhanced in situ performance monitoring of limiting fuel temperatures, self-shielded neutron flux, and potentially spatially distributed strain within complex SiC components if sensors can be successfully embedded during CVI. This work describes (1) methods for embedding sensors; (2) thermodynamic analyses and material compatibility testing for identifying sensors capable of surviving high temperatures and exposure to hydrogen and hydrogen chloride during CVI; and (3) nuclear applications for embedded sensors, including potential failure modes during fabrication and during reactor operation. Molybdenum-sheathed thermocouples were successfully embedded in a complex SiC component, whereas niobium-sheathed high-temperature irradiation-resistant thermocouples started to drift as soon as the reactant gases were introduced and ultimately failed during CVI due to severe constrained expansion, potentially resulting from niobium hydride formation in the low-temperature region of the CVI system. Optical fibers were successfully embedded in SiC, but further work is needed to protect the fragile fiber leads after their protective coatings are removed during CVI.

thermocouples↗
DOE OSTI2021

Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation

Ferritic/martensitic (F/M) steels are being considered as potential structural materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the most promising candidates. In this study, two conventional and two composition engineered HT9 alloys were irradiated using 3.5 MeV Fe 2+ up to 600 peak displacement-per-atom (dpa) at 450 °C. Void swelling and microstructure evolution were characterized for each alloy and compared. The two conventional HT9 alloys (INL and ACO3) showed similar void swelling behavior due to their similar elemental composition and processing conditions. The INL HT9 exhibited a maximum of 2.4% swelling and the ACO3 HT9 showed a maximum of 2.8% swelling at 342 and 393 average local dpa, respectively. On the other hand, the two-composition engineered HT9 alloys with varying N contents (10 ppm for low N and 440 ppm for high N) showed disparate swelling behavior. The low N HT9 exhibited a maximum of 4.6% swelling, while the high N HT9 showed a maximum of 0.7% swelling at 342 average local dpa. Changes in the N content also affected Ni/Si rich G-phase formation. The low N HT9 showed a larger size and lower density of G-phase precipitates compared with the high N HT9 after 600 peak dpa irradiation. This study compares the void swelling behavior of the ion irradiated four current HT9 alloys to extremely high doses, with the void swelling data from neutron irradiated HT9 alloys. Furthermore, the comparison lends critical insights into how well these current alloys can withstand high neutron fluxes in future reactors, especially since the low N and high N HT9 alloys have never been exposed to such high doses before.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Ab initio study of tungsten-based alloys under fusion power-plant conditions

Tungsten (W) is considered a leading candidate for structural and functional materials in future fusion energy devices. The most attractive properties of tungsten for magnetic and inertial fusion energy reactors are its high melting point, high thermal conductivity, low sputtering yield, and low long-term disposal radioactive footprint. However, tungsten also presents a very low fracture toughness, primarily associated with intergranular failure and bulk plasticity, limiting its applications. In recent years, several families of tungsten-based alloys have been explored to overcome the aforementioned limitations of pure tungsten. These include tungsten-based high-entropy alloys (W-HEAs) and tungsten-based Self-passivating Metal Alloys with Reduced Thermo-oxidation or “SMART alloys” (W-SAs). Given their proximity to the plasma, it is crucial to understand how the exposure of these candidate plasma-facing materials (PFMs) to the neutron fluxes expected in fusion reactors impacts their material behavior over time. In this work, we present a computational approach that combines inventory codes and first-principles DFT electronic structure calculations to understand the behavior of transmuting tungsten-based PFMs. In particular, we calculate the changes in the chemical composition, production uncertainties, the elastic and ductility properties, and the density of states for five tungsten-based PFMs when exposed to EU-DEMO fusion first wall conditions for ten years.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Expanded method for the determination of burnup in nuclear fuels using multiple neodymium isotopes

Established methods for the determination of burnup in nuclear fuels commonly rely on the measurement of 148Nd in the spent fuel in concert with known fission product yields to determine the atom percent of fissions in the fuel. This isotope of neodymium is used for various reasons, including chemical and radioactive stability, ease of measurement, and low rates of formation and destruction due to neutron flux apart from fission. However, careful calculation of effective cumulative fission yields and correction factors for (n,γ) capture reactions allows for additional stable and long-lived isotopes of neodymium to be used to provide additional independent measurements of burnup, reducing statistical uncertainty. This method was developed and successfully applied to measure the burnup of compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The mean burnup measured using 143Nd, 145Nd, 146Nd, 148Nd, and 150Nd was statistically observed to be the same as that measured using 148Nd alone, but the statistical uncertainty in the measurement was reduced by a factor of 2, providing a tighter confidence interval in the final results.

Helmreich, Grant [ORNL] (ORCID:0000000330464394)↗
DOE OSTI2023

Dose and dose rate effects on the microstructural and mechanical stability of long-range ordered precipitates in Inconel 718

Nickel-based alloys offer exceptional mechanical properties and corrosion resistance for various applications in extreme environments such as high neutron flux, temperature, and stress. Precipitates, resulting from thermal treatments, increase the strength and creep resistance of these alloys. Here, heavy ion irradiation is shown to destabilize the γ' and γ" precipitates in Inconel 718 nickel-based alloy by disordering and dissolution processes. Transmission electron microscopy and atom probe tomography reveal that the disordering and dissolution rates of the precipitates are proportional to both irradiation dose and dose rate. Irradiation induced disordering occurs at a higher rate than the dissolution. From analyzing the hardening contributions, microscale mechanical results suggest radiation-induced disordering is responsible for the majority of the softening response at low dose. In addition, low dose radiation-induced disordering results in strain localization and is believed to be the primary cause for the drastic degradation of uniform elongation and toughness in irradiated Inconel 718.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Nuclear Data Sheets for A=64

The evaluated experimental data are presented and evaluated for 13 known nuclides of mass 64 (Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se). For each nuclide, the best values combining all available data are recommended for spectroscopic properties. No excited states have been identified in 64 Ti, 64 As, and 64 Se. Only one excited state in 64 V as an isomer, and three in 64 Cr have been identified. Data for excited states in 64 Mn, 64 Fe, and 64 Co remain limited. 64 Ni, 64 Cu, and 64 Zn are the most studied nuclides through various reactions and decays, followed by 64 Ga and 64 Ge. In the opinion of the evaluators, there are several incomplete or discrepant aspects of the high-spin portion of the level scheme for 64 Zn above ≈5 MeV excitation which need to be resolved in further experiments using large γ-detector arrays. The decay schemes of 64 Ti β - , 64 As ε and 64 Se ε are unknown, while very little information is available for 64 V β - and 64 Cr β - . The decay schemes of 64 Mn β - , 64 Fe β - , 64 Co β - and 64Ge ε are somewhat better known, but still considered incomplete by evaluators. The decay scheme of 64 Ga ε decay is known in detail, however there is the possibility of additional levels populated above 4713 keV, as the Q(ε) value is 7171 keV. The β - , β + and ε decay modes of 64 Cu are well known, as this radionuclide is of great importance in applications, for example, as a dosimeter for neutron flux determination in nuclear reactors, and as a radiopharmaceutical for positron emission tomography (PET). This work supersedes earlier full evaluations of A=64 by 2007Si04, 1996Si12, 1991Si03, 1979Ha35 and 1974Au04.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

A structural model of the long-term degradation of the concrete biological shield

The concrete biological shield (CBS) of light water reactors is exposed to high neutron radiation dose in the long term, which may lead to the degradation of the concrete’s mechanical properties. Given the important shielding role of the CBS, it is necessary to investigate the irradiation effects at the structural scale and provide estimates of the damage extent from the wall’s inner surface to study potential license renewals. For this purpose, we developed a mechanical model accounting for radiation-induced expansion, creep, and damage in concrete using the Grizzly finite element code, informed by ex-core neutron flux calculations using the VERA tool. The model was applied to a 3D CBS structure represented by the CBS wall, a steel liner, reinforcement bars, and a concrete base mat and evaluated damage at 40, 60, and 80 years of operation. The VERA model predicted a maximum fluence of approximately 2 x 10 19 ncm -2 at 80 years of operation. The results showed that damage is highest at the inner surface of the CBS wall and gradually decreases with depth. It extends beyond the rebar after 60 years and reaches a depth of approximately 12 cm at 80 years.

42 ENGINEERING↗
DOE OSTI2023

Fluid structural interaction simulation of the MP-1 plate performance irradiated in the advanced test reactor

Due to the high neutron fluxes they generate and increased heat transfer performance, plate type fuels are used in the U.S. high-performance research reactors. During irradiation, a significant amount of fission energy (i.e., ~200 MeV per atom) is released by the U-235 chain reaction that is carried away by the coolant. Through thermal hydraulic analysis, the coolant’s heat transfer capability is investigated to ensure that the desired fuel temperature can be maintained. At high temperatures, the plate undergoes elastic/plastic deformation, creep, and swelling as a result of both the temperature gradients and the fission gas production within the fuel. These effects are studied via fuel performance analysis. When the plate deformation is small enough that the coolant flow remains relatively unchanged, conducting these two types of analyses independently will suffice. But at high fuel burnups, the swollen plates may encroach into the coolant channels that separate the fuel plates from each other and cause these channels to narrow. Large reductions in channel gap size imperil cooling performance, causing fuel temperatures to rise. Further, if the plate deformation is asymmetric, the fuel centerline will shift toward one side of the channel, causing an uneven reduction in coolability. In addition, a boehmite (oxide) layer will, over time, grow on the plate surface, further obstructing the heat transfer process. To precisely predict fuel temperatures/deformation, a complete coupled analysis that considers coolant flow, heat transfer, oxide growth, elastic/plastic deformation, creep, and swelling is needed; however, this type of analysis method is not available in the literature. To fill the gap, this research developed a fluid structure interaction (FSI) approach to the fuel plate analysis, then successfully applied it to the Mini-Plate (MP)-1 experiment, which was irradiated in the Advanced Test Reactor (ATR) for both one and two cycles. The complete analysis coupled STAR-CCM+, a computational fluid dynamic (CFD) software for calculating flow, with Abaqus, a finite element analysis code for calculating plate deformation. Improvements in the results were found when comparing the fully coupled analysis to the independently conducted analyses but they were not significant due to the miniature size of the plates and the relatively short irradiation time. In the future, the fully coupled approach presented herein will be applied to full-size fuel plates with longer irradiation cycles once additional experiments become available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Evaluation of the spatial self-shielding impact for TRISO-based nuclear fuel depletion

Reactor physics analyses of nuclear cores with nuclear fuel concepts containing tristructural isotropic (TRISO) particles, such as pebbles or compact fuel elements, rely on various degrees of simplification to keep these highly heterogeneous problems computationally tractable. One such limitation regards the level of spatial discretization employed during burnup calculations, where traditionally only a limited number of spatial zones are modeled at the full core level and assume that the spectrum is constant within the fuel elements and TRISO particles in this depletion zone. This type of assumption neglects the impact of spatial self-shielding effect within the kernels (microscale level) as well as within the compact or pebbles (mesoscale level). Furthermore, the Monte Carlo code Serpent 2 contains many relevant features for efficiently modeling this type of geometry, including a collision-based domain decomposition intended for very large burnup calculations, which we leveraged for this work to quantify the impact of capturing neutron flux variations occurring at the micro- and mesoscale level on a series of high-temperature gas-cooled reactor fuel element depletion problems. While spatial self-shielding is observed at both scales, with differences from a volume-averaged burnup of ±7% within the kernels and ±2% between TRISO particles within the fuel element, the conjugated effect on nuclide inventories and multiplication factor are negligible, hence confirming that assuming a single average spectrum value may be sufficient for most applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Assessment of ITER radiation environment during the remote-handling operation of In-Vessel components with D1SUNED

During ITER operational life, a remote-handled cask will be used to transfer In-Vessel components to the Hot Cell for maintenance, storage and decommissioning purposes. Due to the distribution of penetrations for system allocation in the facility, the radiation field of each transfer operation presents a high spatial variability; all operations must be studied independently for workers and electronics protection. In this paper, we present a fully representative approach to describe the radiation environment during the complete remote-handling scenario of In-Vessel components in the ITER facility. The impact of all relevant radiation sources during different stages of the operation is addressed. As-built structures and 2020 baseline designs are considered to produce the most detailed neutronics model of the Tokamak Complex, the 400,000-tonne civil structure hosting the tokamak, up to date. Novel capabilities of the D1SUNED code have allowed to compute the integral dose, the dose rate and the photon-induced neutron flux of both moving and static radiation sources. Time bins are included in the simulations to compute the dose rate caused by In-Vessel components at all positions along the transfer. The time evolution of the dose rate is built in video format with a 1-m resolution, especially valuable for hot-spots identification.

42 ENGINEERING↗
DOE OSTI2023

Thermal Cross Section Measurements At The RPI LINAC

Recently, a cold moderator was designed and developed for use at the Rensselaer Polytechnic Institute Linear Accelerator (RPI LINAC). This cold moderator proved to easily and safely couple to an existing neutron producing target, while enhancing neutron flux below 0.02 eV by up to a factor of 8 by cooling polyethylene down to 29 K. This cold moderator capability allowed for significantly improved counting statistics below 0.02 eV not previously possible due to a poor signal to background ratio. Additionally, testing was performed to characterize the energy resolution of the new cold moderator system and found the system easily capable of resolving resonances in Ta-181 at 4 and 10 eV, while also clearly resolving the Bragg edges found in Be metal below 0.01 eV. Furthermore, following the design and development of a cold polyethylene moderator, a series of thermal total cross section measurements were performed for polyethylene, polystyrene, Plexiglas and yttrium hydride in the thermal region. These measurements serve to help validate thermal scattering law (TSL) evaluations in the 0.0005 – 1 eV energy range. For polyethylene and polystyrene, two sets of experiments were performed – one with the Enhanced Thermal Target (ETT) and another with the ETT plus the new cold moderator capability (ETTC). The yttrium hydride and Plexiglas measurements were only performed with the ETTC. The measurements for polyethylene help to validate the data processing methodology when using the ETTC, while extending the measured range of polyethylene down to 0.0005 eV. Two different Plexiglas, Plexiglas G and Plexiglas G-UVT, and two different concentrations of yttrium hydride, H/Y = 1.85 and 1.68, were measured. Overall, all materials had generally good agreement with their ENDF/B-VIII.0 TSL evaluations, though some discrepancies were noticed. In the case of the yttrium hydride, the high energy oscillations in the hydrogen cross section and the low energy Bragg edges in the yttrium cross section were clearly seen. These measurements represent the first total cross section measurements that encompass the entire thermal region from 0.0005 – 1 eV for polystyrene and yttrium hydride.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Density functional theory calculation of helium segregation and decohesion effect in W110/Ni111 interphase boundary

Density functional theory (DFT) is employed to study the preferential distribution and decohesion effect of He in a W-NiFe composite consisting of W particles embedded in an Ni-based solid solution matrix. A slab containing {110}<100>W//{111}<110>Ni interface is employed as a surrogate model for the W-NiFe system. Firstly, the fracture energy of the W/Ni interphase boundary (IB) (4.37 J/m 2 ) is higher than the cleavage energy of Ni{111} (3.82 J/m^2) and lower than the cleavage energy of W{110} (6.60 J/m 2 ). The comparison shows that the cohesion of the IB is stronger than the Ni{111} planes that are away from the IB. However, the cohesion between the Ni{111} planes adjacent to the IB is found to be the weakest in this system, with a cleavage energy of 3.11 J/m 2 . Subsequently, formation energy of He is calculated in the Ni slab, W slab, and various interstitial sites in the IB. The DFT calculations show that He is significantly more stable in Ni than in W, by about 1.75 eV. Interestingly, He does not prefer to segregate at the IB as compared to bulk Ni. Nevertheless, it prefers to segregate to the region between the Ni{111} planes adjacent to the IB, and decreases the cohesion of the already weakest region. Based on an estimated amount of He gas production in 5 years under first wall neutron irradiation (neutron flux of 1.04x10 15 n/cm 2 /s), the He will decrease the cleavage energy of the weakest region by 21.2%, 15.4%, and 12.2% at 800, 1000, and 1200 C, respectively.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2021

Developing “inverted-corona” fusion targets as high-fluence neutron sources

We present experimental studies of inverted-corona targets as neutron sources at the OMEGA Laser Facility and the National Ignition Facility (NIF). Laser beams are directed onto the inner walls of a capsule via laser-entrance holes (LEHs), heating the target interior to fusion conditions. The fusion fuel is provided either as a wall liner, e.g., deuterated plastic (CD), or as a gas fill, e.g., D 2 gas. Such targets are robust to low-mode drive asymmetries, allowing for single-sided laser drive. On OMEGA, 1.8-mm-diameter targets with either a 10-μm CD liner or up to 2 atm of D 2 -gas fill were driven with up to 18 kJ of laser energy in a 1-ns square pulse. Neutron yields of up to 1.5 × 1010 generally followed expected trends with fill pressure or laser energy, although the data imply some mix of the CH wall into the fusion fuel for either design. Comparable performance was observed with single-sided (1x LEH) or double-sided (2x LEH) drive. NIF experiments tested the platform at scaled up dimensions and energies, combining a 15-μm CD liner and a 3-atm D2-gas fill in a 4.5-mm diameter target, laser-driven with up to 330 kJ. Neutron yields up to 2.6 × 10 12 were measured, exceeding the scaled yield expectation from the OMEGA data. The observed energy scaling on the NIF implies that the neutron production is gas dominated, suggesting a performance boost from using deuterium–tritium (DT) gas. In this work, we estimate that neutron yields exceeding 10 14 should be readily achievable using a modest laser drive of ~300 kJ with a DT fill.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

PIONEER, a high-resolution single-crystal polarized neutron diffractometer

PIONEER is a high Q-resolution, single-crystal, polarized neutron diffractometer at the Second Target Station (STS), Oak Ridge National Laboratory. It will provide the unprecedented capability of measuring tiny crystals (0.001 mm 3 , i.e., x-ray diffraction size), ultra-thin films (10 nm thickness), and weak structural and magnetic transitions. PIONEER benefits from the increased peak brightness of STS cold-neutron sources and uses advanced Montel mirrors that are able to deliver a focused beam with a high brilliance transfer, a homogeneous profile, and a low background. Monte Carlo simulations suggest that the optimized instrument has a high theoretical peak brilliance of 2.9 × 10 12 n cm -2 sr -1 Å -1 s -1 at 2.5 Å at the sample position, within a 5 × 5 mm 2 region and a ±0.3° divergence range. The moderator-to-sample distance is 60 m, providing a nominal wavelength band of 4.3 Å with a wavelength resolution better than 0.2% in the wavelength range of 1.0–6.0 Å. PIONEER is capable of characterizing large-scale periodic structures up to 200 Å. With a sample-to-detector distance of 0.8 m, PIONEER accommodates various sample environments, including low/high temperature, high pressure, and high magnetic/electric field. Additionally, a large cylindrical detector array (4.0 sr) with a radial collimator is planned to suppress the background scattering from sample environments. Bottom detector banks provide an additional 0.4 sr coverage or can be removed if needed to accommodate special sample environments. We present virtual experimental results to demonstrate the scientific performance of PIONEER in measuring tiny samples.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

Plasma window performance and scaling for an accelerator-based neutron source

A “plasma window” was developed for use with a SHINE accelerator-based neutron source. In this work, the design of the plasma window is presented along with results demonstrating successful operation in deuterium over a range of aperture diameters (5–12 mm), gas flow rates (3.5 to 12.5 slm), and input powers (10–55 kW). Here, an empirically determined relationship between the aperture diameter, deuterium gas flow rate, input power, and plasma window pressure differential is also presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

MHD stability trends and improved performance of LHD inward-shifted configurations: The role of the neutral beam current drive and thermal plasma density

The aim of the present study is to analyze the effect of the neutral beam current drive (NBCD), thermal plasma density, and NBI operational regime on the stability of pressure gradient-driven modes (PGDM) and Alfvén eigenmodes (AE) in LHD inward-shifted configurations. The stabilization of n/m = 1/2 PGDM (n toroidal mode and m poloidal mode) is observed in the discharge 167 800 during the co-NBCD phase. The iota profile evolution measured by motional stark effect diagnostic may indicate the iota profile up-shift caused by the co-NBCD can induce a non-resonant transition of the rational surface 1/2 before the mode stabilization. The evolution of the iota profile and continuum gaps in the discharge 167 805 during the ctr-NBCD phase leads to the stabilization of the AE, caused by the narrowing of the continuum gap as the iota profile down-shift. Opposite stability trends are identified for PGDM and AE stability with respect to the thermal plasma density. A larger thermal plasma density (larger thermal β) further enhances PGDM although the continuum gaps are narrower leading to configurations with stable AEs. The linear stability of AEs is analyzed using the gyro-fluid FAR3d code to reproduce the AE stability trends observed in the experiments with respect to the NBCD and thermal plasma density. The analysis of hypothetical scenarios dedicated to study different NBI operational regimes with respect to EP energy, and β and radial density profiles indicate off-axis NBI operation shows a higher EP β threshold to destabilize AEs compared to on-axis configuration. This is explained by the presence of a TAE gap in the inner plasma region, easily destabilized by an on-axis NBI injection. The control of the NBCD and thermal plasma in the discharge 167 800 shows a transitory stabilization of PGDM and AEs, as well as an improved discharge performance identified by an increment of the neutron fluxes.

Varela, Jacobo↗