Fast Neutron Damage Research Articles

Investigation of irradiated metal of WWER-type reactor internals after 45 years of operation. Part 2. Calculated and experimental determination of the fast neutron fluence and damage dose

Fast neutron flux, fluence and damage dose have been determined when using experimentally measured specific activity of 54 Mn и 58 Co isotopes for metal of samples cut out from elements of pressure vessel internals of Novovoronezh NPP Unit No 3 (18Cr-10Ni-Ti steel, analog of AISI 321 steel). The results have been compared with the values calculated by KATRIN-2.5 computer code.

Newly developed BGO glasses: Synthesis, optical and nuclear radiation shielding properties

In this research paper, we studied the optical and nuclear shielding efficiency of newly developed BGO glasses with the following compositions (in wt%): 32Bi2O3–68GeO2, 42Bi2O3–58GeO2, 47Bi2O3–53GeO2, 52Bi2O–48GeO2, 62Bi2O3–38GeO2. BGO glasses were prepared by traditional melt quenching method. To obtain the band gap values of fabricated BGO glasses, optical absorption spectra were used for evaluation of optical properties. The mass attenuation coefficients (μ/ρ) were achieved for prepared glasses at 0.015–15 MeV photon energies employing MCNPX Monte Carlo code and WinXcom program. Moreover, broad-range of nuclear shielding parameters for gamma ray, neutrons and charged particles such as mass attenuation coefficient, half value layer, effective atomic number, buildup factors, mass stopping powers, projected ranges, fast neutron removal cross sections and damage factors were calculated. The refractive index is calculated from Eopt, As Bi2O3 concentration is enhanced, Eopt is also increased as well as the optical electronegativity and consequently the refractive index. In addition, the results showed that BIGE5 glass sample with highest Bi2O3 contribution has excellent nuclear radiation shielding ability among the other fabricated glass samples.

Quantifying yield behaviour in metals by X-ray nanotomography.

Nanoindentation of engineering materials is commonly used to study, at small length scales, the continuum mechanical properties of elastic modulus and yield strength. However, it is difficult to measure strain hardening via nanoindentation. Strain hardening, which describes the increase in strength with plastic deformation, affects fracture toughness and ductility, and is an important engineering material property. The problem is that the load-displacement data of a single nanoindentation do not provide a unique solution for the material’s plastic properties, which can be described by its stress-strain behaviour. Three-dimensional mapping of the displacement field beneath the indentation provides additional information that can overcome this difficulty. We have applied digital volume correlation of X-ray nano-tomographs of a nanoindentation to measure the sub-surface displacement field and so obtain the plastic properties of a nano-structured oxide dispersion strengthened steel. This steel has potential applications in advanced nuclear energy systems, and this novel method could characterise samples where proton irradiation of the surface simulates the effects of fast neutron damage, since facilities do not yet exist that can replicate this damage in bulk materials.

Models of coefficient of thermal expansion (CTE) for Gilsocarbon graphites irradiated in inert and oxidising environments

This paper presents the development and validation of an empirical model of radiation effects on coefficient of thermal expansion (CTE) for the Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The combined irradiation and oxidation model is based in part on a new model of fast neutron damage in inert environment. The new inert model shows an increase to an “upper shelf” irradiated CTE value at very low dose, then CTE values decrease with increasing dose following a hyperbolic tangent function. The effect of the actual exposure in AGRs is modelled by shifting the inert model in both dose and CTE directions to agree with the CTE measurements on material trepanned from moderator bricks in operating AGRs. The shift in the inert model that is needed to match the trepanned data varies significantly by reactor. The new model predicts randomly-selected validation data that were not used in model fitting as well as it fits the calibration data.

A model of Young’s modulus for Gilsocarbon graphites irradiated in oxidising environments

This paper presents the development and validation of an empirical model of combined fast neutron damage and radiolytic oxidation effects on Young’s modulus for the Gilsocarbon graphites used in advanced gas-cooled reactors. The combined irradiation and oxidation model is based on a model of fast neutron damage in an inert environment and the degree of radiolytic oxidation as measured by weight loss. The change in Young’s modulus with exposure was found to vary from one power reactor to another and with the relative contributions of oxidation and fast neutron damage in data from materials test reactors. The model was calibrated to data from material trepanned from advanced gas-cooled reactor moderator bricks after varying operating times. The model also predicts materials test reactor data that extend beyond the range of currently-available trepanned data. The model is convenient to use because it is an analytic function that eliminates the interpolation on temperature and dose that is used in an earlier model.

Models of bending strength for Gilsocarbon graphites irradiated in inert and oxidising environments

This paper presents the development and validation of an empirical model of fast neutron damage and radiolytic oxidation effects on bending strength for the moulded Gilsocarbon graphites used in Advanced Gas-cooled Reactors (AGRs). The inert environment model is based on evidence of essentially constant strength as fast neutron dose increases in inert environment. The model of combined irradiation and oxidation calibrates that constant along with an exponential function representing the degree of radiolytic oxidation as measured by weight loss. The change in strength with exposure was found to vary from one AGR station to another. The model was calibrated to data on material trepanned from AGR moderator bricks after varying operating times.

Effects due to a Pu-C source on a HPGe detector and the corresponding neutron shielding

A gamma spectrum of a Pu-C source is measured using a p-type HPGe detector, whose three peaks (full energy, single-escape and double-escape peak) can be used as a calibration source for the beam energy measurement system of BEPCII. The effect of fast neutron damage on the energy resolution of the HPGe detector is studied, which indicates that the energy resolution begins to deteriorate when the detector is subject to 2×107 n/cm2 fast neutrons. The neutron damage mechanism and detector repair methods are reviewed. The Monte Carlo simulation technique is utilized to study the shielding of the HPGe detector from the fast neutron radiation damage, which is of great significance for the future commissioning of the beam energy measurement system.

Non-contact characterisation of conductivity gradient in isotropic polycrystalline graphite using inductance spectroscopy measurements

The two main graphite ageing processes in advanced gas-cooled reactors (AGRs) are fast neutron damage and radiolytic graphite oxidation. These processes change the properties of the graphite core and lead to a reduction in the graphite density, ie a weight loss. Electrical conductivity measurement of the graphite material is among the possible methods for estimating the weight loss and hence provides a measure of the graphite ageing. The inductance spectroscopy technique using eddy current sensors has been implemented in a wide variety of NDT applications, since it enables the possibility of extracting the electrical parameters variation from eddy current measurements and examining more fully the internal material structure. In this paper we present a non-contact eddy current method for determining non-destructively the electrical conductivity gradient through 100 mm-thick graphite sections using inductive spectroscopic measurements collected by a gradiometer. The method employs the commercial finite element software COMSOL to generate the data from a model of a real profile distribution. This data is translated to the profile conductivity of the graphite test object via the solution of an inverse problem. The linear Tikhonov regularisation method and the non-linear regularised Gauss Newton technique are employed in the solution of the inverse problem. During the optimisation, the forward problem is evaluated using COMSOL. Numerical optimisation tests have been carried out for different cases of step conductivity profiles. Initial results using simulated data show that representative estimates of the conductivity profile of the modelled graphite sections have been obtained.

Development of a Young’s modulus model for Gilsocarbon graphites irradiated in inert environments

Dimensional and material property changes caused by fast neutron damage, radiolytic oxidation or a combination of both have been measured over many years on the Gilsocarbon graphites. The data have been gathered together in a comprehensive database and are being used to develop dose–damage relationships of irradiated graphite material properties for use in irradiated graphite component stress analyses up to a fast neutron dose of ∼200 × 10 20 n/cm 2 equivalent DIDO nickel dose (EDND) in the temperature range ∼300–650 °C. This paper covers development and validation of an empirical model for Young’s modulus of Gilsocarbon graphite when irradiated in an inert environment. The new model provides a good fit over the range of the primary modelling variables: dose, irradiation temperature, and graphite group. The proposed model is convenient to use in component stress analysis, as it features an analytic function of temperature and dose that eliminates the need in the existing model for interpolation on those variables.

Toward fullerene-based radiopharmaceuticals: high-yield neutron activation of endohedral [formula omitted] metallofullerenes

C 60 and endohedral 165 Ho metallofullerenes were irradiated under various neutron flux conditions to maximize the activity of surviving 166 Ho metallofullerenes. C 60 survived irradiation to a fairly high extent (>80%) at fluences of <10 18 n cm −2 (75.8% thermal), but fast-neutron damage (or equivalent oxidative decomposition) progressed rapidly at higher fluences. Under identical conditions the metallofullerenes were degraded more quickly and extensively than was C 60, and decomposition proceeded primarily by fast-neutron damage instead of anticipated (n, γ)-metal recoil. Irradiation under more thermal neutron flux conditions (96.2% and 99.8%) significantly improved metallofullerene survival (20–30%), and degradation was demonstrated to proceed through the metal-recoil pathway.

Irradiation growth kinetics in 235U-doped zirconium

The in situ irradiation growth kinetics of two specimens of Zr-0.25 at.% Al doped with 0.06 at.% 235U: specimen RX (fully recrystallized) irradiated at 533 K in the CNEA-RA1 at a dose rate of 3.3 × 10−8 dpa s−1, and specimen CWSR (50% cold-worked and stress relieved), irradiated at 618 K in the CNEA-RA3 at a dose rate of 6.6 × 10−7 dpa s−1, are presented. The experimental method, based on the continuous deflection measurement and on the use of fissionable doped material to enhance the fast neutron damage, allows measurement of the growth phenomenon from the early stages of the irradiation. The transient or Stage I results of the irradiation growth behaviour obtained are in good agreement with results shown in the literature.

Role of disordered regions in fast-neutron damage of HPGe detectors

Published data on resolution degradation of high-purity germanium (HPGe) detectors after irradiation with fast neutrons are reviewed and the possible role of point defects is quantified. Resolution degradation is between one and two orders of magnitude too severe to be attributed to the point defects present. Disordered regions must dominate resolution degradation until the detector has been sufficiently annealed that the pre-irradiation resolution is almost recovered. In capacitance spectroscopy the corresponding emission from these disordered regions has presented problems either in identification or interpretation.

Bulk Defects and Radiation Damage in Detector Grade Silicon

ABSTRACTThe importance of bulk defects in Si to the performance of Si radiation detectors is discussed and the current state of knowledge about deep level defects, including those induced by radiation damage, is briefly reviewed. The importance and origins of the fluctuations in the spatial distribution of the shallow point defects which determine the uncompensated net impurity density are discussed and information on this problem in FZ silicon, multipass FZ silicon, neutron transmutation doped Si, and radiation damaged Si is presented and compared to what should be expected on the basis of simple modeling. A new model for radiation damage induced changes in the net uncompensated impurity density is reviewed and compared to experimental data on fast neutron damage in Si.

Effect of fast neutron damage on the performance of a high resolution HPGe coaxial detector

The effect of fast neutron damage on the energy resolution and the full energy peak efficiency (FEPE) of a 157 cm3 coaxial HPGe detector is studied. When the detector is subjected to approximately 5*108 fast neutrons, the energy resolution deteriorates from 1.85 keV to 11.65 keV at 1332 keV. The FEPE decreases by approximately 28% at 80.997 keV and approximately 8% at 1332 keV with a dose of approximately 4*108 fast neutrons. With the same dose of fast neutrons the total detection efficiency decreases by approximately 13%, and the peak to Compton ratio (for 1408.03 keV 152Eu peak) reduces to 11.7 from its original value of 70.6.

Potential of Fast Neutron Damage Regions in Heavily Doped Semiconductors

AbstractThe potential and field of impurity—defect damage regions are calculated, employing a model which is earlier used to explain changes in properties of heavily doped silicon irradiated by neutrons. The change in initial material properties and the effect of the energy of the irradiation particles are analyzed.

Fast neutron damage in silicon detectors

Radiation effects of fast neutrons have been measured in silicon detectors of varying resistivity irradiated to ≈ 10 11 n/cm 2 over periods of weeks. The principal damage effect is increased leakage current due to generation of carriers from defect levels in the depletion region. Damage and leakage current constants have been established for detector resistivities between 10 and 27 000 ω cm and lie in the range of (0.7–2) × 10 7 s/cm 2 ( K) for PuBe neutrons. A slight increase in K was observed for higher resistivities which translates into somewhat improved radiation hardness. A fit of these data was attempted to a two-level recombination formulation of the damage constant.

An NMR and electrical assessment of heavy doping of indium antimonide via the neutron transmutation technique

A series of InSb bars were doped with Sn donors by subjecting them to a thermal neutron flux. Reactor locations with different thermal/fast-neutron flux ratios were used, permitting some assessment of the type of damage emanating from fast-neutron exposure. Various figures of merit for the quality of the sample, such as the total electron concentration n, electrical mobility mu and the nuclear spin-lattice relaxation rate 1/T1, are discussed for samples with reactor exposure times ranging up to several hours. The study attempts to quantify the advantages to be gained from sample irradiation in reactor locations with favourable thermal/fast-neutron flux ratios. The indications are that fast-neutron damage has a detrimental effect on the induced electron concentration.

Elevated temperature tensile properties of irradiated 20/25/Nb stainless steel fuel pin cladding at low and high strain rates

Tensile specimens were prepared from 20/25/Nb stainless steel fuel pin cladding irradiated in a Commercial Advanced Gas-cooled Reactor (CAGR) at temperatures in the range 622–866 K and integrated fast neutron doses up to 16 × 10 24 n/ m 2. The tests were performed in air at temperatures in the range 298–873 K at strain rates from 2 × 10 −5 s −1 to 7.2 s −1. The tensile properties varied with irradiation temperature, test temperature and strain rate. At lower irradiation temperature, strengthening produced by fast neutron damage was accompanied by reduced elongation. Strengthening was also observed at higher irradiation temperatures, possibly due to precipitation phenomena. The maximum irradiation embrittlement was observed in tests at 873 K at low strain rates between 2 × 10 −4 s −1 and 2 × 10 −5 s −1. The failure mode of embrittled specimens irradiated at higher temperatures was characterized by prematurely ruptured ductile fibres, rather than by intergranular cracking.

Fast neutron damage in [formula omitted

Fast (14 MeV) neutron damage in titanium monocarbide (TiC 1−x) was examined by measuring low-temperature electrical resistivity and magnetic susceptibility. Both physical properties changed remarkably with neutron damage. The damage process seemed to depend on the nonstoichiometry (the amount of the carbon vacancies), as had been revealed in an “in-situ” resistivity measurement at liquid He temperature. On the other hand, it is of interest that both properties (without damage) are considerably affected by the nonstoichiometry.

Magnetization studies of YBa 2Cu 3O 7−x irradiated by fast neutrons

Studies of the effect of fast neutron damage on the magnetic hysteresis of YBa 2Cu 3O 7−x ceramic samples subjected to fluence of neutrons of 2∗10 16 n/cm 2 up to 6∗10 17 n/cm 2 have been performed. irradiation up to dose of 1∗10 17 did not cause any change in the critical temperature. However it causes a strong increase of the magnetic hysteresis which is presumably connected with the creation of defects. The critical current density at 77 K in H = 10 k0e for the sample irradiated with the dose 1∗10 17 n/cm 2 was estimated to be 520 A/cm 2 as compared to 29 A/cm 2 for the reference non-irradiated sample, non-irradiated sample.