Helium Gas Production Research Articles

Cost-Effective Thermomechanical Processing of Nanostructured Ferritic Alloys: Microstructure and Mechanical Properties Investigation.

Nanostructured ferritic alloys (NFAs), such as oxide-dispersion strengthened (ODS) alloys, play a vital role in advanced fission and fusion reactors, offering superior properties when incorporating nanoparticles under irradiation. Despite their importance, the high cost of mass-producing NFAs through mechanical milling presents a challenge. This study delves into the microstructure-mechanical property correlations of three NFAs produced using a novel, cost-effective approach combining severe plastic deformation (SPD) with the continuous thermomechanical processing (CTMP) method. Analysis using scanning electron microscopy (SEM)-electron backscatter diffraction (EBSD) revealed nano-grain structures and phases, while scanning transmission electron microscopy (STEM)-energy dispersive X-ray spectroscopy (EDS) quantified the size and density of Ti-N, Y-O, and Cr-O fine particles. Atom probe tomography (APT) further confirmed the absence of finer Y-O particles and characterized the chemical composition of the particles, suggesting possible nitride dispersion strengthening. Correlation of microstructure and mechanical testing results revealed that CTMP alloys, despite having lower nanoparticle densities, exhibit strength and ductility comparable to mechanically milled ODS alloys, likely due to their fine grain structure. However, higher nanoparticle densities may be necessary to prevent cavity swelling under high-temperature irradiation and helium gas production. Further enhancements in uniform nanoparticle distribution and increased sink strength are recommended to mitigate cavity swelling, advancing their suitability for nuclear applications.

Importance of cross reaction covariance data for user applications

The characterization of the uncertainty in radiation damage metrics presents many challenges. This paper examines the current approaches to characterizing radiation damage metrics such as hydrogen and helium gas production, material heating, trapped charge in microelectronics, and lattice displacement damage. Critical uncertainty aspects go beyond just the material cross sections and involve the consideration of energy-dependent cross reaction correlations, the recoil ion energy spectrum, and models used for the partitioning of the recoil ion energy into various forms of energy deposition. This paper starts with a review of terminology and then examines the current approaches in the characterization of uncertainty in radiation damage metrics for several applications. The major deficiencies in the uncertainty of the damage metric characterization are also identified.

Activation and transmutation of tungsten boride shields in a spherical tokamak

The FISPACT-II code is used to compute the levels of activation and transmutation of tungsten borides for shielding the central high temperature superconductor core of a spherical tokamak fusion power plant during operations at 200 MW fusion power for 30 years and after shutting down for 10 years. The materials considered were W2B, WB, W2B5 and WB4 along with a sintered borocarbide B0.329C0.074Cr0.024Fe0.274W0.299, monolithic W and WC. Calculations were made within shields composed of each material, for five reactor major radii from 1400 to 2200 mm, and for six 10B isotope concentrations and at five positions across the shield. The isotopic production and decay in each shield is detailed. The activation of boride materials is lower than for either W or WC and is lowest of all for W2B5. While isotopes from tungsten largely decay within 3 years of shut-down, those from boron have a much longer decay life. An acceptable 70% of the absorbing 10B isotope will remain after 30 years of operations behind the first wall for a 1400 mm radius tokamak. Gaseous production is problematic in boride shields, where 4He in particular is produced in quantities 3 orders of magnitude higher than in W or WC shields. The FISPACT-II displacements per atom (dpa) tend to increase with boron content, although they decrease with increased 10B isotopic content. The dpa ranges of boride shields tend to lie between those of W and WC. Overall, the results confirm that the favourable fusion reaction shielding properties of W2B5 are not seriously challenged by its irradiation and transmutation properties, although helium gas production could be a challenge to its thermal and mechanical properties.

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.

Nuclear performance analyses of the HCLL breeder blanket for a fusion power reactor

This paper summarizes results of a neutronic study on the performance of the helium-cooled lithium–lead (HCLL) blanket for a fusion power reactor. The analysis is based on three-dimensional radiation transport calculations with the Monte Carlo code MCNP-4C. The neutronic model based on the PPCS reactor model AB developed by FZK has been used in the analysis. Nuclear performance parameters such as the tritium breeding ratio and the energy deposition rate in the blanket and shield and their radial distributions have been evaluated. The radiation-induced displacement and the helium gas production in the steel structure has been calculated. It was shown that the blanket configuration considered provides TBR greater than the required design value. The gaps between blanket modules was shown to affect the global TBR only insignificantly. The radiation damage assessment has indicated that the HTS could not be designed as a life-time component of the reactor for the anticipated 40 years of plant operation.

Effects of spectral shifting in an inertial confinement fusion system

Abstract The main objective is to study the effects of spectral shifting in an inertial confinement system for kT/shot energy regime on the breeding performance for tritium and for high quality fissile fuel. A protective liquid droplet jet zone of 2 m thickness is used as coolant, energy carrier, and breeder. Flibe as the main constituent is mixed with increased mole-fractions of heavy metal salt (ThF4 or UF4) starting by 2 moles% up to 12 moles%. Spectrum softening within the inertial confinement system reduces the tritium production ratio (TBR) in the protective coolant to a lower level than unity. However, additional tritium production in the 6Li2DT zone of the system increases TBR to values above unity and allows a continuous operation of the power plant with a self-sustained fusion fuel supply. By modest fusion fuel burn efficiencies (40 to 60 %) and with a few mol.% of heavy metal salt in the coolant in form of ThF4 or % UF4, a satisfactory TBR of > 1.05 can be realized. In addition to that, excess fissile fuel of extremely high isotopic purity with a rate of ∼ 1000 kg/year of 233U or 239Pu can be produced. Radiation damage through atomic displacements and helium gas production after a plant operation period of 30 years is very low, namely dpa < 1 and He < 2 ppm, respectively.

Monte Carlo analysis of helium production in the ITER shielding blanket module

In order to examine the shielding performances of the inboard blanket module in the International Thermonuclear Experimental Reactor (ITER), shielding calculations have been carried out using a three-dimensional Monte Carlo method. The impact of radiation streaming through the front access holes and gaps between adjacent blanket modules on the helium gas production in the branch pipe weld locations and back plate have been estimated. The three-dimensional model represents an 18° sector of the overall torus region and includes the vacuum vessel, inboard blanket and back plate, plasma region, and outboard reflecting medium. And it includes the 1 m high inboard mid-plane module and the 20 mm wide gaps between adjacent modules. From the calculated results for the reference design, it has been found that the helium production at the plug of the branch pipe is four to five times higher than the design goal of 1 appm for a neutron fluence of 0.9 MW a m −2 at the inboard mid-plane first wall. Also, it has been found that the helium production at the back plate behind the horizontal gap is about three times higher than the design goal. In the reference design, the stainless steel (SS):H 2O composition in the blanket module is 80:20%. Shielding calculations also have been carried out for the SS:H 2O composition of 70:30, 60:40, 50:50 and 40:60%. From the evaluated results for their design, it has been found that the dependence of helium production on the SS:H 2O composition in the blanket module is small at the branch pipe. Altering the steel–water ratio to reduce the amount of steel and increasing the thickness by >170 mm will reduce helium production to satisfy the design goal and not have a significant impact on weight limitations imposed by remote maintenance handling limitations. Also based on the calculated results, about 200 mm thick shields such as a key structure in the vertical gap are suggested to be installed in the horizontal gap as well to reduce the helium production at the back plate and to satisfy the design goal.

Neutronic and photonic analysis of the single box water-cooled lithium lead blanket for a DEMO reactor

The water-cooled Pb–17Li demonstration plant (DEMO) breeding blanket line was selected in 1995 as one of the two EU lines to be further developed in the next decade. In this paper the results of a neutronic and photonic analysis of the `single box' concept is presented. A full three-dimensional model, including the whole assembly and many of the DEMO reactor components, has been developed, together with a three-dimensional neutron source. A tritium breeding ratio (TBR) value of 1.16, with no ports and a Li 6 enrichment of 90%, has been obtained and a further analysis has been performed to determine Li 6 enrichment that would still ensure tritium breeding self-sufficiency. Selected power densities, calculated for the different materials and zones, are also presented. Material damage through displacements per atom and helium gas production has been investigated. Some shielding capability considerations are presented too.

Neutronic Calculations for a Magnetic Fusion Energy Reactor with Liquid Protection for the First Wall

Liquids may be used between the magnetic confined fusion plasma and the first wall of the plasma chamber to reduce the material damage through displacements per atom (dpa) and helium gas production. This could extend the lifetime of the first wall in a magnetic fusion energy (MFE) reactor to a plant lifetime of ~30 yr.Neutronic calculations are carried out in S16P3 approximation for a typical HYLIFE-II blanket geometry, an inertial fusion energy (IFE) reactor design. This provides a comparison of the damage data between compressed and uncompressed targets, for IFE and MFE applications, respectively, by using Flibe (Li2BeF4), natural lithium, and Li17Pb83 eutectic as both coolant and wall protection. In the consideration of mainline design criteria, including sufficient tritium breeding ratio (TBR = 1.1), material protection (dpa < 100 and He < 500 parts per million by atom in 30 yr of operation), and shallow burial index, coolant zone thickness values are found to be 60 cm for Flibe, 171 cm for natural lithium, and 158 cm for Li17Pb83 with Type 304 stainless steel (SS-304) as structural material.Material damage investigations are extended to structural materials made of SiC and graphite for the same blanket to obtain waste material suitable for shallow burial after decommissioning of the power plant.The dpa values and helium production rates in graphite are comparable to those in SS-304. However, they are higher in SiC than in SS-304 and graphite.The average neutron heating density in the external 1.6-mm-thick SS-304 shell of the investigated blanket beyond the SiO2 insulation foam decreases rapidly with increasing thickness of the Flibe coolant. With DR = 60 and 80 cm, it becomes only 594 and 95 μW/cm3, respectively. The design limit for heat generation density in superconducting coils for magnetic fusion is 80 μW/cm3. A very important result of this work is that a blanket with liquid-curtain protection would not require extra shielding for superconducting coils around the fusion plasma chamber. This could result in an important simplification of the design.

Radiation Damage in Liquid-Protected First-Wall Materials for IFE-Reactors

Material damage through displacements per atom (DPA) and helium gas production, as well as the tritium breeding and energy absorption in an IFE (Inertial Fusion Energy) reactor chamber have been in...

Determination of (n, charged particle) reaction cross sections for frt-relevant materials

Abstract Helium gas production cross sections for elemental nickel and molybdenum, deduced from the isotopic data, are reported. Cross sections for 58Ni (n,d) 57Co and 10B (n,t)2α reactions were measured. Excitation function of triton emission reaction on aluminium was determined via activation in diverse neutron fields and unfolding code calculations.

The Sensitivity of the First Wall Radiation Damage to Fusion Reactor Blanket Composition

The atomic displacement and hydrogen and helium gas production rates in a 0.01-m-thick Type 316 stainless-steel first wall have been calculated as a function of the composition of a 0.5-m-thick fusion reactor blanket. The atomic displacement rates range from a low value of 8.4 dpa/yr for an empty blanket to a high value of 22.6 dpa/yr for a lead-filled blanket. Hydrogen gas production rates vary from 423 to 536 appm/yr, and the helium gas production rates vary from 134 to 159 appm/yr for the empty and carbon-filled blankets, respectively.

Monte Carlo Analysis of the Effects of a Blanket-Shield Penetration on the Performance of a Tokamak Fusion Reactor

Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm/sup 2/ rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively.