Technique for Radioactivity Measurement in Drum Package Waste by Using Scattered Gamma-Rays

An evaluation method for the radionuclide content in low level drum package waste by using scattered γ-rays has been investigated. Gamma-rays are counted by a detector after passing through a collimator having appropriate vertical and horizontal openings both for the segmented scan and for the average geometrical efficiency being almost independent of the source position. The attenuation of unscattered γ-rays is estimated from the intensity ratio of scattered rays to unscattered ones of the nuclide emitting the highest γ-ray energy in the drum. Satisfactory results were obtained in numerical simulations and basic experiments carried out using test equipment with a pure Ge detector and a rectangular collimator, and sample 200l drums, containing sealed ⁶⁰Co and ¹³⁷Cs sources, and having a uniform density of 2.2 g/cm³. The error of the evaluated content, which depended slightly on the activity distribution, was a maximum of about ±30%.     

Determination of radiation attenuation coefficients of heavyweight- and normal-weight concretes containing colemanite and barite for 0.663 MeV γ-rays

Accurate measurements have been made to determine radiation transmission of concretes produced with barite, colemanite and normal aggregate by using beam transmission method for 0.663 MeV γ-rays energy of ¹³⁷Cs radioactive isotopes by using NaI(Tl) scintillation detector. Linear and mass attenuation coefficients of thirteen heavy- and four normal-weight concretes were calculated. It was determined that the linear attenuation coefficient (μ, cm⁻¹) decreased with colemanite concentration and increased with barite concentration in both type of the concretes. Mass attenuation coefficient values of our concretes were compared with the values proposed by the United States National Institute of Standards and Technology (NIST).     

NXcom – A program for calculating attenuation coefficients of fast neutrons and gamma-rays

This work is concerned with a construction and use of NXcom computer program for calculating the removal and attenuation coefficients of transmitted fast neutrons and γ-rays, respectively, through mixtures, composites, concretes and compounds. The program uses only one input data file for neutrons and γ-rays calculations. For γ-ray attenuation, the program predictions were tested by comparing them with the well-known WinXcom program results and an excellent agreement was noticed. Also, it has been used for calculating the values of macroscopic effective removal cross-sections Σ_R (cm⁻¹) for five new published polyamide and anhydride composites designed for shielding mixed neutron and γ-rays. The obtained values for Σ_R using the program and the reported attenuation thicknesses which were based on the Monte Carlo N-Particle (MCNP) code showed the same trend. The NXcom program can be used as a preliminary effective tool for testing the shielding material against fast neutrons and γ-rays.     

Radiation shielding of concretes containing different lime/silica ratios

The shielding of γ-rays and fast neutrons by concrete has been studied for concretes containing different lime/silica ratios. Calculations were carried out for six different concrete samples. The total mass attenuation coefficients (μ/ρ, cm² g⁻¹) have been computed at photon energies of 1 keV to 100 GeV using the personal computer software package WinXCom. Also the macroscopic effective fast neutron removal cross-sections (Σ_R, cm⁻¹) have been calculated using MERCSF-N program and the removal cross-section database for all required elements. The obtained results showed that the lime/silica ratio of concrete has significant and insignificant effects on μ/ρ and Σ_R values, respectively.     

Determination of γ-spectrometric parameters for Na2CO3 and H2O in Na2CO3-water solution

The mass attenuation coefficients of Na₂CO₃ solved in water was measured for Na₂CO₃ contents up to 0.191 g of salt in cm³ of water at the energy of 0.662 MeV emitted from ¹³⁷Cs. On the basis of the mixture rule the mass attenuation coefficients were calculated for the solutions at each salt content. Comparison between the measured and calculated values revealed, in general, those experimental values of linear and mass attenuation coefficients are smaller than the calculated ones. The total atomic cross-sections and effective atomic numbers were obtained for the mass attenuation coefficients. Least square method has been applied to determine the mass attenuation coefficients, atomic cross-sections, and the effective atomic numbers of Na₂CO₃ and water, separately, in the solution and compared with the corresponding calculated values.     

Gamma-Ray Spectrometric Determination of Isotopic Ratios of Plutonium

Abstract

Plutonium samples (3–30 μg) of various isotopic compositions were measured with a high-resolution Ge(Li) detector. The γ-rays counted are the 43.5 keV of ²³⁸Pu, the 38.7, 51.6 and 129.3 keV of ²³⁹Pu, the 45.2 keV of ²⁴⁰Pu, and the 148.6 keV of ²⁴¹Pu. To estimate the relative photopeak efficiencies of the γ-rays, an empirical curve was derived from measurements of the three γ-rays of ²³⁹Pu adopted as internal standard. Data treatment is described in detail as well as the method of correction to account for the activities of ²³⁷U and ²⁴¹Am inherently accumulating in the plutonium sample. The branching ratios of the γ-rays of plutonium were also evaluated by comparing the γ-ray spectrometric results with mass-spectrometric data.     

Measurement of Gamma-Ray Production Cross Sections of Iron for Incident Neutron Energies between 6 and 33 MeV

Measurement of differential γ-ray production cross sections, i.e. (n, x γ) cross sections, of Fe was made for neutron energies from 6 to 33 MeV. Neutrons used in the experiment were white neutrons produced with (p, n) reactions by 35 MeV protons using a thick Be target. The neutron energy was analyzed by the time-of-flight method and bunched into 3 MeV wide energy bins, for each of which the spectrum of secondary γ-rays produced in an Fe sample was measured by a BGO scintillator at an angle of 144° to the neutron beam direction.

The obtained (n, xγ) cross sections agreed well with other data and the evaluated data file of ENDF/B-IV at neutron energies below 15 MeV where data were existing. The JENDL-3 file overestimated the γ-ray spectra at γ-ray energies of 3 to 7 MeV. The present work newly provided the data in the neutron energy range above 20 MeV. The GNASH calculation made by Young reproduced the measured data fairly well even at these higher energies.     

Precise Measurement of the Gamma-ray Emission Probabilities for 116m In with Half-life of 54.15 min

The γ-ray emission probabilities for ^116mIn with half-life of 54.15 min have been measured by a 4πβ(pc)-γ(HPGe) coincidence apparatus using a live-timed two-dimensional data acquisition system. Eleven γ-ray peaks of ^116mIn were recognised and the uncertainties of the measured emission probabilities for principal γ-rays were found to be less than 1%. On the other hand, the relative intensities for 18 energy γ-rays were determined with improved uncertainties by an ordinary γ-ray spectrometer. The decay parameters were determined with uncertainties of approximately 1%using these values and the evaluated values of weak γ-ray intensities and the internal conversion coefficients.     

A Coaxial Ge(Li) Gamma-Ray Spectrometer

A single-open-end coaxial type Ge(Li) γ-ray detector with an effective volume of 8.7cm³ and a capacitance of 35 pF was made by Li-drifting for 440 hr at an applied voltage of 140 V at 40°C in thermomodule-controlled constant-temperature water coolant. The FWHM energy resolution with an E810F preamplifier was 4.2 and 8.0 keV for γ-rays of 122 and 1,173 keV respectively. The pulse height distributions of γ-rays from an irradiated natural U sample and ¹⁰⁵Rh isotope were measured.     

Proton elastic scattering and proton induced γ-ray emission cross-sections on Na from 2 to 5 MeV

Differential cross-sections for proton elastic scattering on sodium and for γ-ray emission from the reactions ²³Na(p,p′γ)²³Na (E_γ = 440 keV and E_γ = 1636 keV) and ²³Na(p,α′γ)²⁰Ne (E_γ = 1634 keV) were measured for proton energies from 2.2 to 5.2 MeV using a 63 μg/cm² NaBr target evaporated on a self-supporting thin C film.The γ-rays were detected by a 38% relative efficiency Ge detector placed at an angle of 135° with respect to the beam direction, while the backscattered protons were collected by a Si surface barrier detector placed at a scattering angle of 150°. Absolute differential cross-sections were obtained with an overall uncertainty estimated to be better than ±6.0% for elastic scattering and ±12% for γ-ray emission, at all the beam energies.To provide a convincing test of the overall validity of the measured elastic scattering cross-section, thick target benchmark experiments at several proton energies are presented.     

Anticoincidence-Shielded Ge(Li) Gamma-Ray Spectrometer for Marine Environmental Radionuclide Analysis

An anti-Compton γ-ray spectrometer has been developed, which possesses a number of characteristics particularly adapted to direct measurement of low-level complex mixtures of radionuclides in marine environmental samples.

The detector assembly consists of a 42.7 cm³ coaxial Ge(Li) detector and a well-type 6″diam. × 4″ Nal(Tl) detector for the anticoincidence shield. The sample to be measured is placed on the Ge(Li) detector inside the well of the Nal(Tl) crystal.

Suppression of the Compton background is particularly marked at the Compton edge when the assembly is operated in anticoincidence mode. The anticoincidence shield reduces the ¹³⁷Cs Compton edge by a factor of 6.6 to provide a peak-to-Compton edge ratio of 210: 1 without reducing the full-energy peak efficiency by more than 2%. Similarly, for the 1,332.5 keV ⁶⁰Co γ-rays the Compton edge is reduced by a factor of 7 to give a peak-to-Compton ratio of 85:1, though in this case the peak efficiency is reduced by a factor of 1.8 due to the cascade decay.

This spectrometer has permitted the detection of a few pCi of fallout nuclides such as ¹³⁷Cs and ⁹⁵Zr, contained in 30 g environmental wet samples counted at intervals of 1,000 min.     

Variation of Intensity of Environmental Gamma-Rays Measured with Ge(Li) Spectrometer

The variation of the intensity of environmental γ-rays has been studied with a Ge(Li) spectrometer. The 12 γ-rays were selected to be examined and typical characteristics of variations were obtained for each γ-rays. The intensity of γ-rays of ²¹⁴Pb and ²¹¹Bi which are ²²²Rn daughter nuclei increases on rainy days to the severalfold value of the average in the case of no precipitation. These variations were observed to be roughly proportional to rainfall rate in a sequence of precipitation. The γ-rays intensity of nuclides of the thorium series and ¹³⁷Cs is nearly constant within about 3% regardless of the weather. Its variation is about equal to the statistical fluctuation of the present measurements. The mean diurnal variations of f-ray intensity were obtained on days without precipitation. The intensity of ²¹⁴Pb and ²¹⁴Bi j-rays fluctuates nearly sinusoidally with an amplitude of about 6% of the average value and the maximum in the early morning.     

Mass attenuation coefficients,effective atomic numbers and electron densities of undoped and differently doped GaAs and InP crystals

The total mass attenuation coefficients (μ/ρ), for GaAs, GaAs (semi-insulating; S-I) GaAs:Si (N⁺), GaAs:Zn, InP:Fe, InP:Fe–As, InP:S and InP:Zn crystals were measured at 22.1, 25.0, 59.5 and 88.0 keV photon energies. The samples were irradiated with ¹⁰⁹Cd and ²⁴¹Am radioactive point sources using transmission arrangement. The X- and γ-rays were counted by a Si (Li) detector with resolution of 160 eV at 5.9 keV. Total atomic and electronic cross-sections (σ_t and σ_e), effective atomic numbers (Z_eff) and electron densities (N_el) were determined using the obtained μ/ρ values for the investigated crystals.     

Environmental Gamma-Ray Exposure Rates Measured by In-Situ Ge(Li) Spectrometer

A conversion function (G(E) function) was determined to obtain an exposure rate directly from a γ-rays spectrum measured by an in-situ Ge(Li) spectrometer. In order to verify the utility of the G(E) function method, three kinds of detector, namely, a portable Ge(Li) detector, a Nal(Tl) detector and an ionization chamber, were used to measure exposure rate due to environmental γ-rays at the same place, and the measured values of exposure rate were compared with each other. Two exposure rate values were obtained from a γ-ray spectrum measured by the Ge(Li) detector, using two different methods, namely, the G(E) function method and the HASL method. The average of exposure rates obtained by four different methods was 4.15 üR/h, and the deviation was 15% of the average.

By applying both the G(E) function method and the HASL method to an environmental γ-ray spectrum, a total exposure rate can be determined directly without information on the distribution of the radionuclides in soil, and individual exposure rates due to ⁴⁰K, ²³⁸U and ²³²Th series and ¹³⁷Cs of the known distribution in soil can be also determined.     

Measurement of 76Se and 78Se (γ, n) Cross Sections

The (γ, n) cross sections of Se isotopes (⁷⁶Se,⁷⁸Se) were measured to supply fundamental data for estimating the inverse reaction cross section, i.e., the ⁷⁹Se(n, γ)⁸⁰Se cross section. The enriched samples and a reference ¹⁹⁷Au sample were irradiated with laser-Compton scattering (LCS) γ-rays. The excitation function of each (γ, n) cross section was determined for the energy range from each near neutron separation energy to the threshold energy of (γ, 2n) reaction. The energy point corresponding to each cross section was deduced using the accurately determined energy distribution of LCS γ-rays. Systematic (γ, n) cross sections for Se isotopes including ⁸⁰Se were compared with those calculated by using a statistical model calculation code TALYS.     

Effects of neutron irradiation on polymer matrix composites at 5 K and at room temperature: II. Degradation of mechanical properties

Neutron irradiation with a low flux of accompanying γ-rays in the Intense Pulsed Neutron Source was carried out at 5 K and at room temperature on four kinds of polymer matrix composites (filler: E-glass or carbon fiber cloth; matrix: epoxy or polyimide resin). The specimen irradiated at 5 K was warmed up to room temperature before the mechanical test was performed at 77 K. The Young's modulus of these composites scarcely decreases even when a total neutron fluence is 3.0 × 10¹⁸n/cm² (2.1 × 10¹⁸n/cm² for E > 0.1 MeV) for the 5 K irradiation and 1.6 × 10¹⁹n/cm² (8.0 f 10¹⁸ n/cm² for E > 0.1 MeV) for the room-temperature irradiation. The ultimate strength, however, decreases significantly at this neutron fluence for all the composites except the carbon/epoxy composite whose initial strength is comparatively low. Comparison of this result with that obtained for ⁶⁰Co γ-ray irradiation demonstrates that the radiation sensitivity of the glass/epoxy and glass/polyimide composites is 1.8–2.6 times higher towards neutrons than γ-rays. As to the irradiation temperature of 5 K and room temperature, no significant influence on the degradation efficiency of the composite strength is observed under the present conditions of mechanical test.     

Calculation of the cross-sections for fast neutrons and gamma-rays in concrete shields

This work is concerned with the theoretical calculation of both the total mass attenuation coefficients (μ/ρ) for gamma rays and the effective removal cross-sections (Σ_R) for fast neutrons. Calculations were carried out for four types of concrete of different densities namely (1) dolomite–sand ρ=2.5 g.cm⁻³, (2) barite–barite ρ=3.49 g.cm⁻³, (3) magnetite–limonite ρ=3.6 g.cm⁻³ and (4) ilmenite–ilmenite ρ=3.69 g cm⁻³. The total mass attenuation coefficients have been calculated at energies from 10 keV to 1 GeV using the XCOM (version 3.1) computer program and cross- section database for elements from Z=1 to 100. Also the effective removal cross-section have been calculated using the elemental composition of the concrete mixes. The calculated values are widely needed and used as a database for radiation shielding design of research reactors, power station and particle accelerators. The results were displayed in tables and figures. Calculated results were compared with those previously measured and a reasonable agreement was found between them.     

Benchmark data for 1.43 MeV γ-rays penetrating Pb,steel, Al and graphite shields

NH₄VO₃ solution was pumped around an in-pile closed circuit to provide a uniform disc source of 1.43 MeV γ-rays from the decay of ⁵²V. Angular and scalar flux spectra are presented for 1.43 γ-rays penetrating shielding slabs of 1.54 m.f.p. Pb, 0.99 and 1.99 m.f.p. steel, 0.79, 1.40 and 2.10 m.f.p. Al and 1.38 m.f.p. of graphite. The spectra have been obtained from the unfolding of scintillation spectrometer outputs using the code radak. Such data can be used for the evaluation of calculational methods.     

CaF2:Tm (TLD-300) thermoluminescent response and glow curve induced by γ-rays and ions

The thermoluminescent response of CaF₂:Tm after exposure to ⁶⁰Co γ-rays at doses from 0.44 Gy to 8.75 kGy and to low fluences (10⁵ ∼ 10⁸ cm⁻²) of 25 and 40 MeV ¹H, 75 and 120 MeV ³He, 180, 300 and 480 MeV ¹²C, 400 MeV ¹⁶O and 800 MeV ²⁰Ne ion beams, spanning a LET interval up to about 500 keV/μm, has been investigated. A careful deconvolution analysis of the glow curve has been performed in order to obtain information for individual peaks. The region of linear response to gammas extends up to ≈1 Gy, while that for ions includes the complete dose interval covered in the study (up to ≈1.3 Gy). The ratio between the high- and low-temperature structures in the glow curve is correlated with radiation quality and dependence on ion identity, besides LET, is strongly suggested by the data. The thermoluminescent efficiency to ion exposure, with respect to irradiation with ⁶⁰Co γ-rays, shows a different dependence on LET for each of the peaks. In general terms, the efficiency reaches a maximum between 1.2 and 1.4 near 6 keV/μm and decreases for higher LET. Peak 3 displays a unique trend, its relative efficiency is always less than 1.0 and shows a strong monotonic inverse dependence with LET.     

Analysis of Transmitted Gamma-Rays by Multiple Scattering Method, (I)

Dose buildup factors and number spectra of γ-rays transmitted through a homogeneous finite slab have been estimated by the multiple scattering method, taking into account scattering — including back scattering — up to the fourth order.

The calculations were performed for 1.0, 3.0, 6.0, 8.0 and 10.0 MeV γ-rays normally incident on lead, iron and water slabs of thicknesses from 1 to 15 mfp.

The results of the above calculations are in good agreement with those from other calculations, such as by Monte Carlo and response matrix methods, especially for heavy shielding materials of practical importance and γ-rays of high incident energy.

Further, a method is proposed with which the contribution to the total dose buildup factor by the γ-rays of the fifth and higher orders of scattering can be estimated approximately. With this method, good agreement was obtained with the dose buildup factor calculated by the Monte Carlo method, even for light shielding materials and γ-rays of low incident energy.