Fusion Reactor Shield Research Articles

A high temperature W2B–W composite for fusion reactor shielding

We have developed a new material for neutron shielding applications where space is restricted. W2B is an excellent attenuator of neutrons and gamma-rays, due to the combined gamma attenuation of W and neutron absorption of B. However, its low fracture toughness (3–4 MPa m1/2) and high melting point (2670 °C) prevent the fabrication of large fully-dense parts with adequate mechanical properties. Here we meet these challenges by combining W2B with a minor fraction (43 vol%) of metallic W. The material was produced by reaction sintering W and BN powders. The mechanical properties under flexural and compressive loading were determined up to 1900 °C. The presence of the ductile metallic W phase enabled a peak flexural strength of ∼850 MPa at 1100 °C, which is a factor of 2–3 higher than typical monolithic transition-metal borides. It also enabled a ductile-brittle transition temperature of ∼1000 °C, which is not observed in monolithic borides. Compression tests showed hardening below ∼1500 °C and significant elongation of the phase domains, which suggest that by forging or rolling, further improvements in ductility may be possible. These results have implications for W2B–W shield design; neutronics performance will likely improve with increased boron content, however this study suggests mechanical properties and manufacturability may degrade.

Effective Atomic Weight, Effective Atomic Numbers and Effective Electron Densities of Hydride and Borohydride Metals for Fusion Reactor Shielding

The effective atomic weight, Aeff, mass attenuation coefficients, μ/ρ, half-value layer, HVL, effective atomic numbers, Zeff and effective electron densities, Neff of hydride and borohydirde metals, with potential shielding applications in fusion reactors have been investigated in the present work. The gamma ray interaction parameters, μ/ρ, HVL, Zeff and Neff were calculated for photon energy range 1 keV–100 GeV. The Aeff was calculated using fast neutron removal cross-sections (ΣR) for neutron energy 2–12 MeV. The effective atomic number using fast neutron ΣR was evaluated for compound/composite materials first time. The compounds or mixture containing low as well high atomic weight elements are required for fast neutron shielding. The Mg (BH4)2, TiH2 and ZrH2 hydride and borohydride metals are found to be the superior shielding materials. This study should be useful for design of effective shielding using hydrides and borohydrides metals in fusion reactors.

Split-screen single-camera stereoscopic PIV application to a turbulent confined swirling layer with free surface

An annular liquid wall jet, or vortex tube, generated by helical injection inside a tube is studied experimentally as a possible means of fusion reactor shielding. The hollow confined vortex/swirling layer exhibits simultaneously all the complexities of swirling turbulence, free surface, droplet formation, bubble entrapment; all posing challenging diagnostic issues. The construction of flow apparatus and the choice of working liquid and seeding particles facilitate unimpeded optical access to the flow field. A split-screen, single-camera stereoscopic particle image velocimetry (SPIV) scheme is employed for flow field characterization. Image calibration and free surface identification issues are discussed. The interference in measurements of laser beam reflection at the interface are identified and discussed. Selected velocity measurements and turbulence statistics are presented at \( \hbox{Re}_{\uplambda}=70\) (Re = 3500 based on mean layer thickness).

Sensitivity and uncertainty analysis performed on a 14-MeV neutron streaming experiment

In the 14 MeV neutron streaming experiment performed at the Frascati Neutron Generator (FNG) a mock-up of a module of a fusion reactor shielding system was irradiated. A void channel with a high aspect ratio (l/ a>10) was penetrating through the shield and a cavity was placed at the end. Several nuclear quantities (activation reaction rates) were measured at different positions. In this work the experiment has been analysed using discrete ordinates ( S N) code DORT and the sensitivity/uncertainty code package SUSD3D. The uncertainty of the neutron flux at two measurement positions (A — inside the cavity and B — 87.6 cm in the shielding block) due to the iron cross-sections was found to be very low (e.g. the for the flux above 10 MeV between ∼0.2% and 3%). The uncertainty analysis is normally useful to explain the discrepancies between the calculated (C) and the measured (E) nuclear quantities, or at least to eliminate some possible sources for the discrepancy and to quantify uncertainty in design calculation. In this case the uncertainty due to the iron cross-sections seems to be too low to explain the discrepancy between C and E.

Effect of Aggregate Specification in Heavy Concrete for Fusion Reactor Shield on Neutron Dose Evaluation

Heavy concrete is one of the most powerful, reliable and inexpensive radiation shield materials. It can be applied to a fusion reactor as an effective and strong shield. However, heavy concrete inherently has heterogeneous structure due to the absorber aggregates which can cause heterogeneity effect on the shielding performance. In the present study, the aggregate-specification dependent heterogeneity effect in neutron dose evaluation has been studied for D-T neutron field using measured leakage neutron spectra from heavy concrete samples with D-T neutrons and the analysis results conducted with the Monte Carlo code MCNP-CFP.From the results, the heterogeneity effect of heavy concrete could be observed in the evaluation of neutron dose equivalent. The effect was reproduced accurately by the MCNP-CFP calculation. The ratio of dose for heterogeneous heavy concrete to that of homogeneous one was mostly less than unity, and the effect became large with increase of the packing fraction, shield thickness and aggregate diameter. The reason was that the spatial distribution of the aggregates usually tends to be biased towards the center of the shield. In such a case, MCNP-CFP becomes a powerful tool to give the exact shielding performance.

Effect of Heterogeneities in Heavy Concrete on Shielding of Fusion Neutrons

Heavy concrete is a promising candidate material for a fusion reactor shield. It provides strong shielding performance, though it has a heterogeneous structure due to random arrangement of the heavy aggregates contained as absorbers. To obtain data for future fusion reactor shield designs, D-T neutron irradiation experiments using both heterogeneous and homogeneous heavy concrete samples were carried out to investigate how much the heterogeneity due to the aggregates affects shield performance. Leakage neutron spectra and reaction rates of activation foils were measured, and they were compared with the analyses by the Monte Carlo code MCNP-CFP. From the comparison of results, the measured heterogeneity effect was well reproduced by MCNP-CFP, though there was a slight disagreement in the thermal region. For a point neutron source, a heterogeneous shield was found to be advantageous compared with a homogeneous one from the standpoint of shielding performance above 1 MeV. This conclusion was exactly opposite to what was anticipated. Analysis of the results confirmed that the effect was strongly associated with the manufacturing process used for the heavy concrete. For thermal neutrons, a homogeneous shield is still regarded to be most preferable. To suppress the heterogeneity effect above 1 MeV, it is necessary to use a sufficiently large heavy concrete shield. Then analysis with a conventional calculation method is feasible except for the thermal neutron region. If a smaller shield is employed, a specialized Monte Carlo code with a heterogeneous treatment like MCNP-CFP should be used for the precise analysis.

Measurement of Neutron Dose on Fusion Reactor Shield Using TLD-300 Phosphors

benchmark experiment was carried out at the 14 MeV Frascati Neutron Generator (FNG) to measure the nuclear heating in an experimental assembly simulating a fusion reactor shield and its superconducting coils. TLD-300 detectors were used allowing the neutron and gamma contribution to the total nuclear heating to be separated by using the two-peak method. The experiment aimed at validating the EFF and FENDL neutron transport cross section files presently available for nuclear heating calculations of fusion reactor shields. The calculations were performed using the Monte Carlo code MCNP. The measured and calculated gamma absorbed dose agree well while differences up to 30% were found for the neutron absorbed dose when the EFF file is used. The discrepancy increases up to 50% when the FENDL file is used to calculate the neutron absorbed dose.

Comments on “Neutron Streaming through Gaps in Fusion Reactor Shielding”

Comments on “Neutron Streaming through Gaps in Fusion Reactor Shielding”

Neutron Streaming Through Gaps in Fusion Reactor Shielding

Calculations have been performed using the Monte Carlo code, MORSE-CG, to determine the neutron streaming through various straight and stepped gaps between radiation shield sectors in the conceptua...

Monte Carlo Calculations of Neutron and Gamma-Ray Energy Spectra for Fusion Reactor Shield Design: Comparison with Experiment

Neutron and gamma-ray energy spectra resulting from the interactions of approx. 14 MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree witin 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra are also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE.

Calculation of neutron and gamma-ray energy spectra for fusion reactor shield design: Comparison with experiment-II

Measured and calculated neutron and gamma-ray energy spectra resulting from the transport of ∼14 MeV neutrons through a 0.30-m-thick lithium hydride slab and through a 0.05-m-thick lead slab followed by 0.30 m of lithium hydride are compared. Also reported are comparisons of the measured and calculated neutron energy spectra behind an 0.80-m-thick assembly comprised of stainless steel type 304 and borated polyethylene. The spatial dependence of the gamma-ray energy deposition rate measured using thermoluminescent detectors is compared with calculated data. The calculated data obtained using two-dimensional radiation transport methods and ENDF/B-IV cross-section data are in good agreement for all of the experimental configurations. Calculated integral neutron energy spectra agree with the measured data within ∼5 to ∼20% depending on neutron energy for the LiH and Pb plus LiH assemblies. The gamma-ray spectra agree within 20% for these slabs. The measured and calculated neutron energy spectra behind the SS-304-borated polyethylene assembly agree within ∼5% except at neutron energies below ∼5 MeV where background radiation influences the measured spectra. The gamma-ray energy deposition rates as a function of depth agree within a factor of two at all detector locations.

Fusion reactor shield optimization in terms of personnel access.

The shield composition of a D-T burning fusion reactor has been optimized to minimize the shutdown dose rate outside the shield. A method of reducing the shutdown dose from a stainless steel structure is devised. First of all the 14 MeV neutron flux must be at-tenuated to reduce the products of the (n, p) reactions such as 58Co and 54Mn. Secondly low energy neutrons penetrating through the shield must be decreased to reduce troublesome (n, γ) reaction products such as 59Fe and 51Cr. As the result the criterion of 2.5 mrem/h one day after shutdown outside the shield is shown to be almost satisfied with 1.3 m thick non breeding blanket/shield, and with 1.5 m thick breeding blanket/shield.

Fusion Reactor Shield Optimization in Terms of Personnel Access

The shield composition of a D-T burning fusion reactor has been optimized to minimize the shutdown dose rate outside the shield. A method of reducing the shutdown dose from a stainless steel structure is devised. First of all the 14 MeV neutron flux must be attenuated to reduce the products of the (n,p) reactions such as 58Co and 54Mn. Secondly low energy neutrons penetrating through the shield must be decreased to reduce troublesome (n,γr)reaction products such as 59Fe and 51Cr. As the result the criterion of 2.5 mrem/h one day after shutdown outside the shield is shown to be almost satisfied with 1.3 m thick non breeding blanket/shield, and with 1.5 m thick breeding blanket/shield.

Calculation of Neutron and Gamma-Ray Energy Spectra for Fusion Reactor Shield Design: Comparison with Experiment

Integral experiments that measure the transport of ∼14-MeV deuterium-tritium (D-T) neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out at the Oak Ridg...