• Journal of Radiation Protection and Research
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• v.33 no.4
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• pp.143-150
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• 2008

In this study, we examined the effect of hydroxyl radical generated by $\gamma$-ray and UV irradiation on shrinkage of vitreous body. Change in gel ratio of vitreous body and change in the properties of its components (collagen, sodium hyaluronate) were analyzed. By comparing these results, the amount of hydroxyl radical, which induces the considerable shrinkage of vitreous body, was evaluated from theoretical calculation based on experimental condition and some reported kinetic parameters. It was concluded that the integrated amount of hydroxyl radical required to liquefy half of the vitreous body (Vitreous body gel ratio = 50%) was estimated as $140\;{\mu}molg^{-1}$ from $\gamma$-ray irradiation experiment. Also, from UV irradiation experiment result, it was confirmed that the effect of hydroxyl radical is larger than that of other reactive species. The causes of shrinkage of vitreous body are supposed as follows, 1) decrease in viscosity by cleavage of glycoside bond in sodium hyaluronate, 2) leaching of collagen from vitreous body and 3) leaching of crosslinked products and scission products of collagen.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.269-274
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• 2022

An artificial neural network (ANN) that identifies radionuclides from low-count gamma spectra of a NaI scintillator is proposed. The ANN was trained and tested using simulated spectra. 14 target nuclides were considered corresponding to the requisite radionuclide library of a radionuclide identification device mentioned in IEC 62327-2017. The network shows an average identification accuracy of 98.63% on the validation dataset, with the gross counts in each spectrum N_c = 100~10000 and the signal to noise ratio SNR = 0.05-1. Most of the false predictions come from nuclides with low branching ratio and/or similar decay energies. If the N_c>1000 and SNR>0.3, which is defined as the minimum identifiable condition, the averaged identification accuracy is 99.87%. Even when the source and the detector are covered with lead bricks and the response function of the detector thus varies, the ANN which was trained using non-shielding spectra still shows high accuracy as long as the minimum identifiable condition is satisfied. Among all the considered nuclides, only the identification accuracy of ²³⁵U is seriously affected by the shielding. Identification of other nuclides shows high accuracy even the shielding condition is changed, which indicates that the ANN has good generalization performance.

• Journal of Radiation Protection and Research
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• v.26 no.4
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• pp.393-398
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• 2001

The non-combustible RI wastes disposed of in hospital every year emit ${\gamma}$-ray or ${\beta}$-ray but their activities are very low to the extent of background. Development of more simple methods is needed because the conventional detection methods are so ineffective and complex. In this study, to solve this problem, detection method using efficiency curve for ${\gamma}$-ray emitting radioactive wastes measurement is proposed and experimental detection efficiency equation is also determined through HPGe's standard specimen measurement. For ${\beta}$-emitting radioisotopes detection, new measurement method using detection efficiency estimated by Monte Carlo simulation and SBD measurements is also proposed. According to the results of this paper, the unknown activity of low-level radioactive wastes without LSC requiring the preparation of standard sample and measurement for standard source detection efficiency could be determined efficiently and simply about ${\pm}17%$ in errors by using the theoretical detection efficiency and the SBD measurement result.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1421-1430
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• 2022

The computation of the solid angle and the detector efficiency is considering to be one of the most important factors during the measuring process for the radioactivity, especially the cylindrical γ-ray NaI(Tl) detectors nowadays have applications in several fields such as industry, hazardous for health, the gamma-ray radiation detectors grow to be the main essential instruments in radiation protection sector. In the present work, a generic numerical simulation method (NSM) for calculating the efficiency of the γ-ray spectrometry setup is established. The formulas are suitable for any type of source-to-detector shape and can be valuable to determine the full-energy peak and the total efficiencies and P/T ratio of cylindrical γ-ray NaI(Tl) detector setup concerning the truncated conical radioactive source. This methodology is based on estimate the path length of γ-ray radiation inside the detector active medium, inside the source itself, and the self-attenuation correction factors, which typically use to correct the sample attenuation of the original geometry source. The calculations can be completed in general by using extra reasonable and complicate analytical and numerical techniques than the standard models; especially the effective solid angle, and the detector efficiency have to be calculated in case of the truncated conical radioactive source studied condition. Moreover, the (NSM) can be used for the straight calculations of the γ-ray detector efficiency after the computation of improvement that need in the case of γ-γ coincidence summing (CS). The (NSM) confirmation of the development created by the efficiency transfer method has been achieved by comparing the results of the measuring truncated conical radioactive source with certified nuclide activities with the γ-ray NaI(Tl) detector, and a good agreement was obtained after corrections of (CS). The methodology can be unlimited to find the theoretical efficiencies and modifications equivalent to any geometry by essential sufficiently the physical selective considered situation.

• Journal of Radiation Protection and Research
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• v.6 no.1
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• pp.8-24
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• 1981

This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute(ANSI) N666. These data are used to calculated the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from $2.5{\times}10^{-8}$ to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoetiergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be a useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions.

• 대한방사선방어학회:학술대회논문집
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• 2009.04a
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• pp.250-251
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• 2009

• Journal of Radiation Protection and Research
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• v.11 no.1
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• pp.77-82
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• 1986

Determination of the relation between the kerma(Kinetic Energy Released in Material) and the absorbed dose is one of the basic problems of dosimetry. Kerma and absorbed dose were measured for 6 MV X-ray from the high energy medical linear accelerator and $^{60}Co$ gamma-ray. The experimental results show that the absorbed dose in the transient equilibrium region practically coincide with the kerma in water and Al for $^{60}Co$. The maximum dose depths were $1.45g/cm^2$ for 6MV X-ray and $0.48g/cm^2\;for\;^{60}Co$ gamma-ray. The ratios of the absorbed dose at maximum build-up to the collision kerma at the surface, ($K^{att}$), were 0.949 for 6MV X-ray and 0.992 for $^{60}Co$ gamma-ray. No difference was found between water and Al when the standard field size was used. This results show that the dependence of $K^{att}$ on the material is very small.

• Journal of Radiation Protection and Research
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• v.36 no.1
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• pp.16-23
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• 2011

Accurate measurement of uranium enrichment is very important in nuclear material accountability. The analysis uncertainty of the uranium enrichment measurement with gamma-ray analysis was studied in the present work. FRAM (Fixed energy Response function Analysis with Multiple efficiencies) code was used to determine the uranium enrichment. If the shield materials were placed between the detector and the sample, the error was measured and analyzed. Measurement time was varied and the dependency of the analysis uncertainty on the measurement time was studied. Transmitted gamma-ray intensities and FWHMs of the peaks in the energy spectrum were measured as the shield thickness was varied. The transmitted gamma-ray intensity follows shape of the exponential function, and the FWHM was almost independent of the shield thickness. The uncertainty of FRAM analysis was studied when the thick shield material was placed between the detector and the sample. Our work could be helpful in analysis of the fissile material in uranium sample.

• Journal of Radiation Protection and Research
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• v.28 no.1
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• pp.19-24
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• 2003

The self-attenuation factor for an $^{198}Au$ sample and the 0.412 MeV gamma-ray penetration ratio in the circular Al-cover of the radiation detector have been determined using an analytical solution and MCNP code. The results show that the self-attenuation factors obtained from the analytical solution coincide with those of MCNP code for all but the Au sample with the relatively larger radius. Then the maximum difference between the two methods appears to be 9 % in the Au sample of 1.5 mm radius. It also is revealed that the analytical solutions of the 0.412 MeV gamma-ray penetration ratio in the Al-cover of 7.62 cm radius are consistent with those of the MCNP code within the standard deviation.

• Plant Biotechnology Reports
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• v.4 no.3
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• pp.237-242
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• 2010

Six parent and their 12 gamma ray-induced somatic flower colour mutants of garden rose were characterized to discriminate the mutants from their respective parents and understanding the genetic diversity using Random amplification of polymorphic DNA (RAPD) markers. Out of 20 primers screened, 14 primers yielded completely identical fragments patterns. The other 7 primers gave highly polymorphic banding patterns among the radiomutants. All the cultivars were identified by using only 7 primers. Moreover, individual mutants were also distinguished by unique RAPD marker bands. Based on the presence or absence of the 48 polymorphic bands, the genetic variations within and among the 18 cultivars were measured. Genetic distance between all 18 cultivars varied from 0.40 to 0.91, as revealed by Jaccard's coefficient matrix. A dendrogram was constructed based on the similarity matrix using the Neighbor Joining Tree method showed three main clusters. The present RAPD analysis can be used not only for estimating genetic diversity present in gamma ray-induced mutants but also for correct identification of mutant/new varieties for their legal protection under plant variety rights.