Helium Generation Research Articles

Ionization and double excitation of helium by short intense laser pulses

We present an ab initio calculation of single ionization, harmonic generation, and double excitation of helium by short (3.8--15.2 fs) moderately strong (peak intensity $1.49\ifmmode\times\else\texttimes\fi{}{10}^{14}{\mathrm{W}/\mathrm{c}\mathrm{m}}^{2})$ laser pulses with wavelengths between 40 and 152 nm. We find excitation to be dominated by resonances with bound and doubly excited states. The processes are nonperturbative. The finite pulse duration leads to resonance broadening and far-off-resonant transitions.

Properties of vanadium-base alloys irradiated in the Dynamic Helium Charging Experiment

One property of vanadium-base alloys that is not well understood in terms of their potential use a fusion reactor structural materials, is the effect of simultaneous generation of helium and neutron damage. In the present Dynamic Helium Charging Experiment (DHCE), helium was produced uniformly in the specimen at linear rates of ≈ 0.4 to 4.2 appm helium/dpa by the decay of tritium during irradiation to 18–31 dpa at 425–600°C in Li-filled capsules in a sodium-cooled fast reactor. This paper presents results of postirradiation examination and tests of microstructure and mechanical properties of V5Ti, V3Ti1Si, V8Cr6Ti, and V4Cr4Ti (the latter alloy has been identified as the most promising candidate vanadium alloy). Effects of helium on tensile strength and ductility were insignificant after irradiation and testing at > 420°C. However, postirradiation ductilities at < 250°C were higher than those of the non-DHCE specimens (< 0.1 appm helium), whereas strengths were lower, indicating that different types of hardening centers are produced during DHCE and non-DHCE irradiation. Ductile—brittle transition behavior of the DHCE specimens was also determined from bend tests and fracture appearance of transmission electron microscopy (TEM) disks and broken tensile specimens. No brittle behavior was observed at temperatures > − 150°C in DHCE specimens. Predominantly brittle-cleavage fracture morphologies were observed only at − 196°C in some specimens that were irradiated to 31 dpa at 425°C during the DHCE. For the helium generation rates in this experiment (≈ 0.4–4.2 appm He/dpa), grain-boundary coalescence of helium microcavities was negligible and intergranular fracture was not observed.

Development of triple fission technique for simulating fusion neutron irradiation

Fusion relevant helium generation can be realized by a technique called ‘triple fission method’. This is essentially an injector-foil technique where the specimen is sandwiched with a pair of foils of fissionable material, e.g. uranium. Energetic alpha particles emitted in a small fraction of fission reactions are injected to adjacent specimens. This technique has been applied for the first time to the irradiation of a reduced activation ferritic steel developed for fusion applications. The results of the first experiment performed in the JMTR is reported in this paper. It is concluded that this technique is useful for simulating helium effects in ferritic steels where other simulation techniques are not easily applicable.

The microstructure of fast neutron irradiated Nimonic PE16

The microstructures of a number of conditions of Nimonic PE16, irradiated in EBR-II at temperatures from 394 to 636°C to displacement doses up to 74 dpa, have been examined. Volume changes, determined from density measurements, were less than 1% in the solution treated and aged (STA) and overaged (OA) conditions. Specimens which were cold worked by 20% or more prior to aging exhibited no significant void swelling. Redistribution of γ′ to point defect sinks occurred, particularly at higher temperatures, resulting in a skeletal γ′ form in both the STA and OA conditions. In cold worked conditions the γ′ precipitates tended to retain a more equi-axed shape. Continuous grain boundary layers of γ′ were produced in STA PE16 at intermediate irradiation temperatures. Grain boundary precipitation of M 6C and Laves phase occurred at higher temperatures and was more prevalent in a high-boron alloy. Mass spectrometry measurements confirmed that the helium generation rate in PE16 was about 1.2 appm per dpa, and assessments of the total helium contents of gas bubbles yielded a value for the bubble surface energy of 3 J m −2. The mechanism of irradiation embrittlement in PE16 is discussed in the light of these observations.

Characterization of nuclear transmutations in materials irradiation test facilities

This study presents a comparison of nuclear transmutation rates for candidate fusion first wall/blanket structural materials in available fission test reactors with those produced in a typical fusion spectrum. The materials analyzed in this study include a vanadium alloy (V4Cr4Ti), a reduced activation martensitic steel (Fe9Cr2WVTa), a high conductivity copper alloy (CuCrZr), and the SiC compound. The fission irradiation facilities considered include the EBR-II (Experimental Breeder Reactor) fast reactor, and two high flux mixed spectrum reactors: HFIR (High Flux Irradiation Reactor) and SM-3 (Russian reactor). The estimated transmutation and displacement per atom (DPA) rates in these test reactors are compared with the calculated transmutation and DPA rates characteristic of a D-T fusion first wall spectrum. In general, past work has shown that the displacement damage produced in these fission reactors can be correlated to displacement damage in a fusion spectrum; however, the generation of helium and hydrogen through threshold reactions (( n,xα ) and ( n,xp )) are much higher in a fusion spectrum. As shown in this study, the compositional changes for several candidate structural materials exposed to a fast fission reactor spectrum are very low, similar to those for a characteristic fusion spectrum. However, the relatively high thermalized spectrum of a mixed spectrum reactor produces transmutation rates quite different from those predicted for a fusion reactor, resulting in substantial differences in the final composition of several candidate alloys after relatively short irradiation time. As examples, the transmutation rates of W, Ta, V, Cu, among others, differ considerably when the irradiation is performed under a mixed spectrum reactor spectrum and under a fusion first wall spectrum. The out-of-core positions in mixed spectrum reactors can be partially shielded against low energy neutrons, e.g. by hafnium, to reduce the transmutation rates, but the displacement rates at these positions are much lower than those for the core positions. Fast reactors (EBR-II) provide the only possibility for obtaining high damage rates without producing significant compositional effects in vanadium alloys, ferritic steels and copper alloys.

Helium content and induced swelling of neutron irradiated beryllium

The effects of neutron irradiation on beryllium in terms of helium generation and induced swelling have been studied. Beryllium samples were irradiated in the Belgian materials testing reactor BR2 at low temperature (40–50 °C) up to 20 000 appm He (for a fission neutron fluence of 5.3 × 10 22 n cm −2). Post-irradiation annealing was performed to study the effect of temperature on the helium behaviour and the induced swelling. Macroscopic examinations of the gas content revealed a linear correlation of 4He content with fast neutron fluence. 4He contents were found to be about 10% higher than their expected formation as calculated from the various formation schemes. The 3He formation was also measured. The 3He calculations have to take into account the continuous transition between 3He and 3H according to the scheme ▪ the latter reaction occurring only when the reactor is in operation. Macroscopic examinations of the swelling by density measurements revealed a linear correlation of low temperature irradiation-induced swelling with He content. This swelling was found to be strongly enhanced after post-irradiation annealing at temperatures above 250 °C. The enhancement of the swelling could be correlated with (He content) and with (annealing temperature). Microscopic examinations on the samples have been started to elucidate the behaviour of He in beryllium. Preliminary results are given.

Influence of helium generation rate and temperature history on mechanical properties of model Fe-Cr-Ni alloys irradiated in FFTF at relatively low displacement rates

In agreement with earlier studies conducted at higher displacement rates, the evolution of mechanical properties at room temperature of model Fe-Cr-Ni alloys irradiated at lower displacement rates in the 59Ni isotopic doping experiment does not appear to be strongly affected by large differences in helium generation rate. This insensitivity to helium/dpa ratio is exhibited during both isothermal and non-isothermal irradiation. The overall behavior of the model alloys used in this study is dominated by the tendency to converge to a saturation strength level that is independent of thermomechanical starting state and helium/dpa ratio, but which is dependent on irradiation temperature and alloy composition.

The influence of 10B addition on the swelling suppression by phosphorus and titanium in Fe-Cr-Ni

Phosphorus and/or titanium modified Fe-16Cr-17Ni austenitic alloys doped with 10B were irradiated in FFTF/MOTA at 684 K for the purpose of investigating the influence of helium generation on the swelling suppression effects by phosphorus and titanium. The increase in void density caused by 10B addition resulted in swelling increase in phosphorus-modified alloys. The close relation between the dislocation density and the void swelling implied the enhancement of dislocation density as a possible mechanism of the swelling suppression by phosphorus addition at the present temperature.

Methods, Methodologies and Formulas for Simplified Neutronics Analyses of Fusion Reactors

The six articles on the simplified neutronics analyses of a fusion reactor published in other journals by the author in 1991-1994 are reviewed. Although the most results were derived under the limitation of materials and models considered and the requried accuracy for the designs, the description may be permissible from the fusion reactor design-oriented point of view. The four simplified blanket engineering methods which have been developed to evaluate the tritium generation in boronized shields, the lithium and beryllium burn-up in blankets and the bismuth and the polonium build-ups in LiPb-bearing blankets, are briefly summarized. The sets of differential equations for the densities of relevant isotopes have been solved using simplified equations. The four shielding methodologies which have been developed to evaluate the fast (E>0.1 MeV) neutron flux, the dose rate, the displacement damage rate and the helium generation rate in in-vessel components of a fusion reactor are also introduced briefly. The abovelisted neutronics responses with the distance from the first wall are described by the exponentially decreasing functions. Attenuation lengths λ required to characterize the attenuation along the thickness of in-vessel composition are obtained for different compositions. The methodology to estimate the total nuclear heating in Toroidal Field (TF) coils is discussed as well. The simplified analytical formulas are also reviewed for 14-MeV and fast neutron fluxes at the exit from a radiation shield with a straight-through diagnostic channel at some distance from the channel orifice. The formulas are derived by fitting the results of the two-dimensional transport calculations. The Simon-Clifford and the Shin formulas are employed in the proposed streaming formulas to treat the 14-MeV and the fast neutron fluxes, respectively, inside the diagnostic channel.

The influence of He/dpa ratio and displacement rate on microstructural evolution: a comparison of theory and experiment

A kinetic model was developed to investigate the influence of the displacement rate and helium generation rate on microstructural evolution in austenitic stainless steels. The model integrates the rate equations describing the evolution of point defects, small point defect clusters, helium-vacancy clusters, and the larger cavity size distribution that is responsible for observable swelling. Cavity (bubble) nucleation is accounted for by the helium-vacancy cluster evolution, while void formation occurs when bubbles grow beyond a critical size in the larger cavity distribution. A series of ion irradiation experiments were used to both calibrate the model and to provide a comparison between model predictions and experimental observations. The experiments involved single and dual-beam irradiations of solution annealed AISI-316 stainless steel at 873 K. The displacement rates were in the range of 2 × 10 −3 to 1 × 10 −2 dpa/s and the helium-to-dpa ratios were in the range of 0 to 50 appm He/dpa. The maximum displacement dose was 25 dpa. The experiments revealed a significant effect of helium on both the dislocation structure and the cavity distribution. The model predictions of helium effects over a broad range of He/dpa ratios and displacement rates were consistent with experimental observations.

Methods, Methodologies and Formulas for Simplified Neutronics Analyses of Fusion Reactors.

The six articles on the simplified neutronics analyses of a fusion reactor published in other journals by the author in 1991-1994 are reviewed. Although the most results were derived under the limitation of materials and models considered and the requried accuracy for the designs, the description may be permissible from the fusion reactor design-oriented point of view. The four simplified blanket engineering methods which have been developed to evaluate the tritium generation in boronized shields, the lithium and beryllium burn-up in blankets and the bismuth and the polonium build-ups in LiPb-bearing blankets, are briefly summarized. The sets of differential equations for the densities of relevant isotopes have been solved using simplified equations. The four shielding methodologies which have been developed to evaluate the fast (E>0.1 MeV) neutron flux, the dose rate, the displacement damage rate and the helium generation rate in in-vessel components of a fusion reactor are also introduced briefly. The abovelisted neutronics responses with the distance from the first wall are described by the exponentially decreasing functions. Attenuation lengths λ required to characterize the attenuation along the thickness of in-vessel composition are obtained for different compositions. The methodology to estimate the total nuclear heating in Toroidal Field (TF) coils is discussed as well. The simplified analytical formulas are also reviewed for 14-MeV and fast neutron fluxes at the exit from a radiation shield with a straight-through diagnostic channel at some distance from the channel orifice. The formulas are derived by fitting the results of the two-dimensional transport calculations. The Simon-Clifford and the Shin formulas are employed in the proposed streaming formulas to treat the 14-MeV and the fast neutron fluxes, respectively, inside the diagnostic channel.

Materials analysis of the TITAN-I reversed-field-pinch fusion power core

The operating conditions of a compact, high-neutron-wall-loading fusion reactor severely limit the choices for structural, shield, insulator, and breeder materials. In particular the response of plasma-facing materials to radiation, thermal and pressure stresses, and their compatibility with coolants are of primary concern. Material selection issues are investigated for the compact, high mass-power-density TITAN-I reactor design study. In this paper the major findings regarding material performance are discussed. The retention of mechanical strength at relatively high temperatures, low thermal stresses, and compatibility with liquid lithium make vanadium-base alloys a promising material for structural components. Based on limited data, the thermal creep behaviour of V3TiISi and V15Cr5Ti alloys is approximated using the modified minimum committment method. In addition, the effects of irradiation and helium generation are superimposed on the creep behavior of V3Ti1Si. Coolant compatibility issues are investigated. The liquid lithium compatibility of the two vanadium alloys, V15Cr5Ti and V3Ti1Si, are compared, and the latter was chosen as the primary structural-material candidate for the liquid-lithium-cooled TITAN-I reactor. Electrically insulating materials, capable of operating at high temperatures are necessary throughout the fusion reactor device. Electrical insulator-material issues of concern include irradiation induced swelling and conductivity. Both issues are investigated and operating temperatures for minimum swelling and dielectric breakdown strength are identified for spinel (MgAI2O4).

Microstructural evolution of neutron-irradiated Fe-Cr-Ni ternary alloys at high helium generation rates

A series of Fe-Cr-Ni ternary alloys were irradiated to 12–14 dpa in the Oak Ridge Research Reactor (ORR). In this experiment, most of the helium atoms were produced by the 58Ni(n, γ) 59Ni(n, α) 56Fe sequential reaction, and, therefore, the helium generation rate was almost proportional to the nickel content, ranging from 27 to 58 appm He/dpa in 20 to 45 wt% Ni alloys. Irradiation temperatures were 330, 400, 500 and 600°C, and the specimen variables were the nickel level, the chromium level and cold-working. The high levels of helium, combined with details of the temperature history, were found to have dominated to microstructural development, sometimes overwhelming and even reversing the action of a given material variable. Microscopy examination showed that there was an unprecedented refinement of cavity microstructure at all temperatures, which accounts for the substantial alterations in both swelling and mechanical properties observed in earlier studies.

A Computational Study of Defects Evolution during Irradiation in Single-Phase Austenitic Alloys

A composite model of the point defect processes and the extended defect evolution in irradiated materials was composed for a theoretical investigation of the response of fusion first wall materials to neutron bombardment. The point defects model calculates the concentration changes of single point defects including a transmutant helium, simple point defect clusters and complex clusters. The extended defect model consists of individual rate theory models of evolution of cavities, Frank faulted loops, network dislocation and other microstructural features. The model was calibrated based on the dual-ion experimental data on an Fe-15Cr-20Ni ternary austenitic model alloy. Using the calibrated model, the effects of helium generation on point defect processes, cavity nucleation, dislocation evolution and mechanical property change were investigated

Linear stability analysis of helium-filled cavities in SiC

In this paper, we present a theoretical analysis of the stability of cavities in SiC under fusion irradiation conditions. Nucleation and growth of helium-filled cavities is modeled by dynamical equations, The growth/shrinkage of a cavity is given in terms of two equations representing the rates of change of the net number of vacancies and helium atoms inside the cavity. The equations describe the drift motion of cavities in a two-dimensional phase under the influence of point defect and helium absorption. First the linear stability of these equations is performed to identify the nature of singular points in phase space. The directions of trajectories around the saddle point are also found. The effects of dislocation density and void density on the nature of the singular points are considered. The effect of helium atom generation rate is also investigated. It is found that in the case of high helium generation, as in SiC under fusion conditions, the critical cavity size is relatively small (about 30 A), and one saddle point is found between two stable critical points. This is shown to correspond to a bimodal cavity distribution.

Helium diffusion and bubble nucleation in the dislocation core during irradiation

The theory of homogeneous nucleation is used to determine the relationship between the diffusion kinetics of helium and the density of helium bubbles both in the matrix and dislocations. The flux of helium atoms reaching a dislocation line is taken to be the effective helium generation rate for the dislocation. In this calculation the effect of the presence of sinks in the matrix is taken into account. Furthermore, the calculation determines the effect of continuously changing flux of helium atoms to the dislocation core due to changes in the sink strength in the grain interior as a function of irradiation time and helium generation rate. The bubble spacing on a dislocation line is then calculated in terms of diatomic nucleation using the condition of nucleation saturation.

Modeling the effects of damage rate and He/dpa ratio on microstructural evolution

The effects of displacement damage rate, helium generation rate and other irradiation conditions on irradiation-induced microstructural changes and swelling in an austenitic stainless steel were investigated by numerical calculations. The computer model developed in this study dynamically integrates rate equations which describe the evolution of point defects, small point defect cluster including cascade vacancy clusters, cavity nucleation and the evolution of the cavity size distribution. The model was calibrated using data obtained from a dual-beam ion irradiation experiment. In this experiment, solution annealed type 316 stainless steel was irradiated by 4 MeV nickel ions with simultaneous implantation of helium ions up to 25 dpa at 873 K. The He/dpa ratios and the displacement damage rates were in the range of 1 to 50 appm He/dpa and 2 × 10 −3 to 1 × 10 −2 dpa/sec, respectively. These experiments showed that helium effects on dislocation evolution significantly influence the nucleation and growth of cavities. The predicted effects of helium over a broad range of damage rates were calculated using the calibrated model.

Helium generation rates in isotopically tailored FeCrNi alloys irradiated in FFTF/MOTA

Three FeCrNi alloys have been doped with 0.4% 59Ni for side-by-side irradiations of doped and undoped materials in order to determine the effects of fusion-relevant levels of helium production on microstructural development and mechanical properties. The alloys were irradiated in three successive cycles of the Materials Open Test Assembly (MOTA) located in the Fast Flux Test Facility (FFTF). Following irradiation, helium levels were measured by isotope dilution mass spectrometry. The highest level of helium achieved in doped alloys was 172 appm at 9.1 dpa for a helium(appm)-to-dpa ratio of 18.9. The overall pattern of predicted helium generation rates in doped and undoped alloys is in good agreement with the helium measurements.

The effect of helium on microstructural evolution and mechanical properties of austenitic steels as determined by spectral tailoring experiments

Fe15Cr xNi alloys irradiated at both low (0.66 to 1.2) and very high (27 to 58) helium/dpa levels exhibit significantly different levels of strengthening due to an unprecedented refinement of cavity microstructure at the very high helium levels. When compounded with the nickel dependence of helium generation, the cavity distribution for some irradiation conditions and alloy compositions can be driven below the critical radius for bubble-to-void conversion, leading to a delay in swelling. The critical radius also appears to be dependent on the nickel level. The refinement may not have resulted from the high helium levels alone, however, but may also have been influenced by differences in displacement rate and temperature history in the two experiments.

The influence of helium on mechanical properties of model austenitic alloys, determined using 59Ni isotopic tailoring and fast reactor irradiation

Tensile testing on model FeCrNi alloys removed from four discharges of the 59Ni isotopic doping experiment in FFTF indicates that helium/dpa ratios typical of fusion reactors do not produce changes in the yield strength or elongation that are significantly different from those at much lower helium generation rates. It also appears that tensile properties approach a saturation level that is dependent only on the final irradiation temperature, but not prior temperature history or thermomechanical starting condition.