Helium Accumulation Research Articles

Density changes in plutonium observed from accelerated aging using Pu-238 enrichment

In support of Stockpile Stewardship activities, accelerated aging tests on a plutonium alloy enriched with 7.3at.% of 238Pu is underway using dilatometry at 35, 50, and 65°C and immersion density measurements of materials stored at 50°C. Changes in density are expected from radiation damage in the lattice and helium in-growth. After 25 equivalent years of aging, the dilatometry data shows that the alloys at 35°C have expanded in volume by 0.11–0.12% and have started to exhibit a near linear expansion behavior primarily caused by the helium accumulation. The average He-to-vacancy ratio from tested specimens was determined to be around 2.55. The model for the lattice damage and helium in-growth accurately represents the volume swelling at 35°C. The density converted from the dilatometry corresponds well to the decreasing density trend of reference plutonium alloys as a function of time.

Zirconolite for Minor Actinide Containment and Alpha Irradiation Resistance

Zirconolite is a potential matrix for the immobilization of the minor actinides: neptunium, curium, americium, and small quantities of unrecyclable plutonium, produced by the reprocessing of the spent fuel.In order to check the incorporation of actinides into the structure, zirconolite ceramic pellets doped with 10 wt% of 239PuO2 were sintered. Characterization by scanning electron microscopy, X-ray diffraction, and X-ray Absorption Near-Edge Structure (XANES) spectroscopy have been done on this material. The microstructure of the pellets is homogeneous, and their relative density is higher than 90% of the theoretical density. XANES spectroscopy shows that Pu is at oxidation state (IV) in this material.To investigate the effects of radiation damage on zirconolite structure, pellets doped with 10 wt% of 238PuO2 were fabricated. 238Pu accelerates the radiation damage relative to the 239Pu because of its much higher specific activity (632 × 109 Bq/g for 238Pu versus 2.2 × 109 Bq/g for 239Pu). Some pellets are stored at ambient, 250 and 500°C. Up to 2.2 × 1018 α·g−1, macroscopic swelling of the samples stored at ambient is ~2.2%/1018 α·g−1, and the microscopic one near 1.3%/1018 α·g−1. Some microcracks are observed on these pellets. The samples started to become amorphous at 2.2 × 1018 α·g−1. The swelling strongly decreases with the storage temperature of the samples.Swelling results are interpreted in terms of alpha radiation damage on the structure; at this time helium accumulation does not appear to have a major role in this process.

Volume changes in δ-plutonium from helium and other decay products

Small changes in volume are observed in plutonium, and several potential mechanisms have been proposed. In this paper, we provide a detailed theoretical analysis of volume changes due to the accumulation of helium and daughter products generated in the radioactive decay of plutonium. It is shown that volume changes in δ-phase plutonium caused by Am, U, and Np are significant and compensate to some degree the swelling from helium bubble formation and growth. Comparison with experimental results obtained so far suggests that the decay products dominate the rate of volumetric change in the long run.

Effects of alpha self-irradiation on actinide-doped spent fuel surrogate matrix

ABSTRACTDuring the initial period after disposal, in which no mass transfer occurs with environmental materials, spent fuel undergoes modifications mainly arising from the effects of α self-irradiation damage. To simulate this aging phenomenon and estimate the consequences on the matrix, an experimental study was carried out in which old UO2 samples doped with short-lived actinides were characterized and analyzed. Changes that occur rapidly at the beginning of the aging process were identified in two lots of actinide-doped samples. Volume and microscopic swelling increased, reaching a maximum relative level of 1.5%. At the same damage level, microscopic swelling remained below 0.9%, suggesting that helium formation and atomic defects have a role in this regard. Similarly, the sample microhardness rapidly increased by up to more than 12%. Characterizing 238PuO2 sources showed at very high integrated and instantaneous doses, irradiation damage and helium accumulation at the grain boundaries resulted in grain decohesion or even powder formation. Helium mobility appeared to be modified by the presence of defects, and local implantation experiments were devised in the CEA's Pierre Süe Laboratory at Saclay to quantify this thermally enhanced diffusion component. The physical condition of the sintered PuO2 pellets after 30 years of storage revealed a purely athermal diffusion mechanism, depending on the sample α activity.

Effect of Helium Accumulation on the Spent Fuel Microstructure

AbstractThe rapid release of activity when water firsts contacts the spent fuel surface in disposal will depend on the pellet microstructure at the arrival time of water in the container. Research performed on spent fuel evolution in a closed system has shown that the evolution of microstructure under disposal conditions should be governed by helium behavior with the cumulated α{decay damage. The evolution of fission gas bubble characteristics under repository conditions has been assessed. In UO2 fuels with a burnup of 47.5 GWd/t, the pressure of fission gas bubbles with the input of helium atoms should not reach the critical bubble pressure, thus micro-cracking in grains is not expected.

Computed Tritium Concentration in a Heavy-Water Reactor

The solutions of a system of equations governing the accumulation of tritium and helium in the heavy water of a reactor are presented for arbitrary initial conditions. The solutions are presented as functions of heavy-water exchange and helium extraction. The concentration change is illustrated for typical values of the thermal-neutron flux density.

Dynamic Simulation of Multiplier Effects of Helium Plasma and Neutron Irradiation on Microstructural Evolution in Tungsten

Plasma-facing and high heat flux materials in a fusion reactor suffer two types of damage: displacement damage caused mainly by high- energy neutrons and surface damage, such as erosion, sputtering, and blistering, caused by hydrogen and helium plasma. Usually, these two kinds of damage are investigated separately. In the present study, multiplier effects of helium plasma and neutron irradiation on microstructural evolution in tungsten were investigated using computer simulations based on a rate theory. Neutron irradiation with 10 � 6 dpas � 1 at 873 K and a helium flux of 10 18 m � 2 � s � 1 with 10-keV energy were used as typical irradiation conditions. The effects of irradiation temperature and defect production rate on the interaction between helium and defects were also investigated. The simulation results revealed the rapid diffusion of helium in tungsten. The helium concentration was saturated in tungsten 0.067 mm thick within 0.01 s at 873 K, and the formation of helium- vacancy clusters was nearly uniform in the matrix except at the surfaces facing towards and away from the irradiation after 0.01 s. This meant that the 10-keV helium plasma could enhance formation of helium-vacancy clusters in the region without damage produced by helium. The concentration of 6He-V clusters reached a very high level (10 � 6 ) even after a 1-s irradiation with a defect production rate of 10 � 6 dpas � 1 at 873 K. The concentration of helium-vacancy clusters decreased with increasing irradiation temperature and decreasing defect production rate. The accumulation of helium and the formation of helium-vacancy clusters depended on not only the vacancy concentration, which was determined by the irradiation temperature and defect production rate, and helium concentrations, but also irradiation time.

Thermal helium desorption behavior in advanced ferritic steels

Thermal helium desorption measurements were performed to investigate the difference in the helium trapping and accumulation behavior among a reduced activation ferritic (RAF) steel and oxide dispersion strengthening (ODS) steels after implantation of He + ions at room temperature. Thermal helium desorption spectra (THDS) were obtained during annealing to 1200 °C at a heating rate of 1 °C/s. The THDS of the ODS steels are very similar to that of the RAF steel, except for the presence of the peak in the temperature range from 800 to 1000 °C, where the α–γ transformation related helium desorption from the γ-phase is considered to occur in the 9Cr-ODS martensitic steels. The fraction of helium desorption becomes larger at higher temperatures, and this trend is increased with the amount of implanted helium. In the 9Cr-ODS steels, the fraction of helium desorption by bubble migration mechanism was smaller than that in the RAF steel. This suggests that the bubble formation was suppressed in the ODS steels. In the 12Cr-ODS steel, the fraction of helium desorption by bubble migration reached more than 90%, suggesting that the trapping capacity of martensite phase in the 9Cr-ODS steel is rather large.

High-beta steady-state research and future directions on the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the Japan Atomic Energy Research Institute Fusion Torus-2 Modified

In the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U), a high-βp ELMy H-mode (high-poloidal-beta high-confinement-mode with edge localized mode) plasma was sustained with βN∼2.7 for 7.4 s. Real-time neoclassical tearing mode (NTM) stabilization system was established and effective NTM suppression by early electron cyclotron (EC) wave injection was demonstrated. High fusion triple product of ni(0)τETi(0)=3.1×1020 keV⋅s⋅m−3 was achieved using the negative-ion based neutral beam current drive with βN∼2.5 and the bootstrap current fraction fBS∼50%. In a hot electron regime, a high electron cyclotron current drive efficiency of 4.2×1018 A/W/m2 was achieved at Te∼21 keV. Innovative current start-up scenario produced a current hole plasma with a very high fBS∼90%. No accumulation of helium and carbon impurities was observed for internal transport barrier (ITB) plasmas. While argon impurity was accumulated, EC injection effectively exhausted it across ITB. In a regime of ELM disappearance, a clear correlation between the ELM frequency and the toroidal velocity at pedestal was observed. In the Japan Atomic Energy Research Institute Fusion Torus-2 Modified (JFT-2M), high beta plasmas were produced with full ferritic inside wall up to βN=3.3, where high recycling steady H-mode discharges were developed up to βNH89P∼6 at ne/nGW∼0.7–1.0 with ITB. JT-60U started long pulse experiment in late 2003 and JFT-2M will conduct wall stabilization experiment in early 2004. The modification of JT-60 to a fully superconducting coil tokamak is regarded as the national centralized tokamak facility program to accomplish the high beta steady-state research in a collisionless regime.

Characterization of neutron field in the experimental fast reactor JOYO for fuel and structural material irradiation test

The experimental fast reactor JOYO has been operated as an irradiation test facility for fast reactor fuel and structural material since 1983 with its MK-II core. During this time, an extensive study was conducted to characterize the neutron field in order to assure the accuracy and reliability of neutron fluence. Neutron flux for a given irradiation test was calculated using a core management code system based on three-dimensional diffusion theory. It was then corrected with the adjusted neutron spectrum by means of the multiple foil activation method. The neutron fluence calculation accuracy in the fuel region was evaluated within a 5% error by comparing the burn-up of spent fuel with the measured values, which had been obtained from their post-irradiation examination. At positions away from the fuel region, the neutron flux distribution was calculated using a two-dimensional transport code. A Monte Carlo code was also used to analyze the detailed neutron flux distribution within an irradiation test subassembly that had a heterogeneous internal structure. With the neutron flux results various irradiation parameters, such as displacement per atom (dpa) and helium production, could be evaluated. A helium accumulation fluence monitor has been developed to measure not only neutron fluence but also helium production. Neutron flux and fluence obtained from the core management calculations were compiled as a database for users’ convenience together with related irradiation information and fuel subassembly material compositions. These data are expected to be widely used in the post-irradiation analysis of fuel and structural material.

Helium accumulation in metals during irradiation – where do we stand?

In the present contribution, the basic ideas pertinent to modelling helium accumulation in metals are reviewed. Topics of earlier work are: Diffusion of atomic He and bubble nucleation under irradiation, low temperature (diffusion controlled) vs. high temperature (dissociation controlled) nucleation, cascade induced He resolution and continuous bubble nucleation at high doses, homogeneous in the bulk vs. heterogeneous nucleation at extended defects; bubble coarsening upon annealing, migration and coalescence vs. Ostwald ripening; bubble state, upper limit vs. ‘equilibrium’ pressure, high pressure equation of state of He, bubble-to-void transformation under a stress or irradiation induced effective vacancy supersaturation. More recent topics are: formation and growth limitation of He platelets (He-filled nano-cracks) in some metals and ceramics, coupled two-component Ostwald ripening of bubble–loop complexes. Possible effects of bubble formation on mechanical properties are briefly addressed: hardening and embrittlement, particularly at high temperature where intergranular fracture is induced by the transformation of bubbles to voids at grain boundaries. Finally, important but unsolved problems are identified and their relevance are briefly discussed.

Helium Accumulation Behavior in Iron Based Model Alloys

Helium desorption from Fe-based model alloys irradiated by energetic helium ions was measured during post-irradiation annealing to investigate the energetics and kinetics of formation and annihilation of helium-related defects. Desorption temperatures were observed to be widely ranged from 450 to 1500 K, indicating that helium is bound to a wide variety of trapping sites such as vacancies and dislocations at various binding states. Such a feature is also observed in fusion ferritic steel. A comparison of helium desorption spectra obtained using Fe, Fe-Cr and Fe-Cr-Ni alloys showed that helium is more strongly trapped in bcc Fe than fcc Fe. It indicates that the long distance migration of helium takes place less frequently in bcc matrix, which may reduce the probability of helium clustering. Fusion ferric steel has a lot of trapping sites for helium such as dislocations, solute atoms, the interface of precipitates, impurities and lath boundaries, and so on, and in addition, it has bct matrix, indicating that most of helium atoms must be dispersed in the matrix and therefore it is difficult for them to cluster as a bubble. This may be a reason for higher helium resistance of the steel.

The effects of grain size and helium accumulation on radiation hardening and loss of ductility of pure copper for ITER application

This paper focuses on the effect of the grain size and high rate of helium accumulation on radiation resistance of pure copper for ITER applications. The paper presents the results of irradiation of three different pure copper grades in the dose range of 0.001–0.1 dpa at 80 °C in the SM-2 reactor. Irradiation at temperatures of 80 °C causes hardening of pure copper. The uniform elongation of samples of all pure copper grades is rather low, the total elongation after irradiation amounts to 10–30%. It is shown that hardening and loss of ductility of pure copper irradiated in the SM-2 to 10 −3–10 −1 dpa at 80 °C are determined mainly by two factors: irradiation dose and grain size. Accumulation of high helium concentrations does not have an essential effect on embrittlement.

Determining the energy parameters of the helium accumulation centers in irradiated boron carbide from thermodesorption spectra

A model for description of the thermodesorption spectra is proposed and used to determine the energy parameters of the centers of helium accumulation in neutron-irradiated boron carbide. It is shown that helium accumulation centers broken at 100–150°C are characterized by a breakage activation energy of 0.23 eV. These centers probably represent clusters or complexes of interstitial helium atoms. The helium accumulation centers broken at 800–1000°C have a breakage activation energy of 0.65–0.75 eV and probably represent helium-vacancy complexes (pores).

Dependence of the energy confinement in the L-and H-modes on the tokamak aspect ratio

In the most advanced tokamak experiments, the H-mode discharges are characterized by an energy confinement time that is higher than in the L-mode by a factor of 1.5–2. For this reason, the H-mode is usually chosen as one of the basic regimes of a tokamak reactor and the presence of internal transport barriers is sometimes assumed as an additional condition. Most present-day tokamaks, which provide the main experimental information, have aspect ratios of 2.6–3.5. In this paper, attention is given to the fact that the enhancement factor (the ratio between the energy confinement times in the H-and L-modes) decreases as the tokamak aspect ratio increases. At aspect ratios ≥5, the confinement times in the L-and H-modes become equal. This result follows from a comparison of the H-and L-mode confinement scalings and is confirmed by experiments in the T-10 tokamak, which has an aspect ratio of 5, and by the results from the international confinement databases. The dependence of the enhancement factor on the aspect ratio can be important for designing steady-state fusion reactors, which are usually characterized by a rather high aspect ratio of 4–8. The possibility of using L-mode discharges with the same confinement time as in the H-mode, but with a lower peripheral temperature and in the absence of edge localized modes (ELMs), can substantially facilitate the operation of the divertor—one of the most stressed construction element of the tokamak reactor. In addition, the L-mode regime does not require overcoming the power threshold for the L-H transition; this circumstance is favorable for extending the range of the working parameters and heating scenarios. The absence of the accumulation of impurities and helium in L-mode discharges may also be important. All of these factors allow one to consider the L-mode as a possible operating regime for high-aspect-ratio tokamak reactors.

Helium Accumulation Behavior in Iron Based Model Alloys

Helium Accumulation Behavior in Iron Based Model Alloys

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

Determination of the helium diffusion coefficient in nuclear waste storage ceramics by a nuclear reaction analysis method

Host matrices for actinide immobilisation will undergo the formation of large helium quantities due to alpha decay. Helium diffusion rate has to be known in order to predict the long-term behaviour of the material, and particularly, the influence of helium accumulation on mechanical properties. A nuclear reaction analysis method, namely the 3He(d, p) 4He reaction, has been used to analyse the evolution of 3He profiles after ion implantations at 1 and 3 MeV in two materials, monoclinic ZrO 2 (as a test material) and Ca 9Nd(PO 4) 5(SiO 4)F 1.5(OH) 0.5 britholite (envisaged for Am and Pu long-term storage). Two data processing methods are used: the classical excitation curve (proton yields versus deuteron energy) and second, the proton energy spectrum for a given deuteron energy. The characteristics of the 3He profiles (depth, width) obtained by both methods are compared to SRIM estimations. Their evolution during subsequent annealings allows an estimation of the helium diffusion rate in the britholite: D ( cm 2/ s)=(2.5±1.5)×10 −4 exp(−(1.07±0.03 eV)/ kT) in the temperature range 200–400 °C, in agreement with previous results on similar materials. Moreover, the shape of the proton energy spectra suggests channeling effects in britholite.

High dose neutron irradiation damage in beryllium as blanket material

The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8×10 22 and 8.0×10 22 cm −2 ( E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70–150 °C and neutron fluence of 2.8×10 22 cm −2 ( E>1 MeV) makes up 0.8–1.5%, at 400 °C and fluence of 8×10 22 cm −2 ( E>1 MeV) – 3.2–5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

Theory for relative strengths of trapping of He+ ions in solid, liquid and gaseous hydrogen

The study of trapping of He+ ion in solid hydrogen is important both as a problem in solid state physics and also as an applied physics problem in the field of muon catalyzed fusion (μCF). In μCF, He+ ion acts as a trap for μ−, interrupting the chain reaction aspect of the catalytic role of μ− in producing fusion of deuteron and triton and of triton and triton in solid hydrogen composed of 2H–3H and 3H–3H molecules, respectively. Using the Hartree–Fock procedure, combined with procedures for including many-body effects, as well as relaxation effects associated with the He+–H2 distances and the adjustment of the H–H separation, we have investigated the trapping of He+ in gaseous and solid state environments. For the former, the environment of He+ is simulated by a single hydrogen molecule and for the solid by clusters appropriately chosen to represent the hexagonal close-packed structure. Our results for the gaseous state indicate that the trapping is rather strong with a binding energy of 8.5 eV, with almost equal binding energy in the linear and triangular configurations with respect to the H–H direction. For the solid, both the likely sites for He+ trapping, namely the tetrahedral and octahedral interstitial sites, are also found to provide deep traps (8.6 eV) of almost equal strength, independent of the orientations of the neighboring molecules, showing that the trapping is not influenced by the orientational disorder in the surrounding hydrogen molecules. Further, the influence of next nearest neighbor hydrogen molecules is found to enhance the trapping energy for He+ substantially, by 0.6 eV, with the incorporation of the third nearest neighbors having a much smaller added effect, demonstrating the convergence of our results with respect to the size of the cluster chosen to simulate the solid. The substantial influence on the He+ trapping energy found for the neighbors beyond the nearest ones provides an explanation of the greater accumulation of helium in the solid state of hydrogen in μCF experiments as compared to the liquid. Suggestions are made regarding the possible reasons for the almost negligible accumulation of helium in the liquid state.