• Journal of Radiation Protection and Research
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• v.43 no.3
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• pp.85-96
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• 2018

Background: A variety of inorganic scintillators have been developed and improved for use in radiation detection and measurement, and in situ gamma-ray spectrometry in the environment remains an important area in nuclear safety. In order to verify the feasibility of promising scintillators in an actual environment, a performance test is necessary to identify gamma-ray peaks and calculate the radioactivity from their net count rates in peaks. Materials and Methods: Among commercially available scintillators, $LaBr_3(Ce)$ scintillators have so far shown the highest energy resolution when detecting and identifying gamma-rays. However, the intrinsic background of this scintillator type affects efficient application to the environment with a relatively low count rate. An algorithm to subtract the intrinsic background was consequently developed, and the in situ calibration factor at 1 m above ground level was calculated from Monte Carlo simulation in order to determine the radioactivity from the measured net count rate. Results and Discussion: The radioactivity of six natural radionuclides in the environment was evaluated from in situ gamma-ray spectrometry using an $LaBr_3(Ce)$ detector. The results were then compared with those of a portable high purity Ge (HPGe) detector with in situ object counting system (ISOCS) software at the same sites. In addition, the radioactive cesium in the ground of Jeju Island, South Korea, was determined with the same assumption of the source distribution between measurements using two detectors. Conclusion: Good agreement between both detectors was achieved in the in situ gamma-ray spectrometry of natural as well as artificial radionuclides in the ground. This means that an $LaBr_3(Ce)$ detector can produce reliable and stable results of radioactivity in the ground from the measured energy spectrum of incident gamma-rays at 1 m above the ground.

• Analytical Science and Technology
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• v.25 no.1
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• pp.14-18
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• 2012

Rn-222 was used as a natural radioactive isotope tracer to evaluate non-aqueous phase liquid(NAPL) contaminated soil and aquifer. In the case of soil sample, Rn-222 concentration was inversely decreased with diesel concentration in the granite soil sample and it was decreased about 30% at the 13% diesel contaminated soil. For evaluating trichloroethylene (TCE) contaminated aquifer, the natural radioisotope Rn-222 was used as naturally occurring partitioning tracer for the approximate localization and semiquantitative assessment of the TCE source zone. Rn-222 was analyzed for the estimation of TCE contamination ranges of the acquifer in the contaminated site at Wonju in Korea.

• Journal of the Korean Society for Nondestructive Testing
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• v.14 no.4
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• pp.228-236
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• 1995

In this paper the method to evaluate shielding deficiency by gamma scanning test was presented and verified theoretically by Monte Carlo code which is one of the best effective method for radiation shielding calculation. The cylindrical shielding model was selected to evaluate shielding deficiency by gamma scanning test. First, the reference shielding according to the design requirement of cask was fabricated specially and reference values were measured with Co-60 source and scintillation detector. As a result with which calculated the reference values, it is shown that maximum deficiency thickness for lead of true cylindrical shielding model was 12mm. To verify this, thickness of lead was calculated by MCNP code and maximum deficiency thickness was 11.6mm. The experimental result obtained by the use of reference shielding was in good agreement with the theoretical result within 4.1%. So, this method can be applied to inspect the shielding ability for great shielding or cask which the radioactive material is used. To perform measurement more exactly, the further work on the development of measuring equipment to display the results on the screen will be required.

• Journal of Korean Society on Water Environment
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• v.22 no.1
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• pp.23-29
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• 2006

Biosorption technology was used to remove hazardous materials from wastewater, herbicide, heavy metals, and radioactive compounds, based on binding capacities of various biological materials. Biosorption process can be explained by two steps; the first step is that target contaminants is in contact with microorganisms and the second is that the adsorbed target contaminants is infiltrated with inner cell through metabolically mediated or physico-chemical pathways of uptake. Until recently, no information is available to explain the definitive mechanism of biosorption. The purpose of this study is to evaluate biosorption capabilities of organic matters using activated sludge and to investigate affecting factors upon biosorption. Over 49% of organic matter could be removed by positive biosorption reaction under anoxic condition within 10 minutes. The biosorption capacities were constant at around 50 mg-COD/mg-MLSS for all batch experiments. As starvation time increased under aerobic or anaerobic conditions, biosorption capacity increased since higher stressed microorganisms by starvation was more brisk. Starvation stress of microorganisms was higher at aerobic condition than anaerobic one. As temperature increased or easily biodegradable carbon sources were used, biosorption capacities increased. Consequently, biosorption can be estimated by biological -adsorbed capability of the bacterial cell-wall and we can achieve the cost-effective and non -residual denitrification with applying biosorption to the bio-reduction of nitrate.

• Journal of the Institute of Electronics Engineers of Korea SP
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• v.46 no.6
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• pp.146-152
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• 2009

The steel-tube is exposed to a radiation from X-ray source. The transmitted radiation is detected by a detector, usually film or more recently an imaging plate(IP) of Computed Radiography(CR). The detected radiation overlaps the region of both sides of the object. The radiographic images reflect the projections of the rays, passing twice through both external and internal tube material. Nonlinear distortion due to the radioactive transmission and geometric disposition also appears on images. In this paper, an analytical approach is presented to achieve image reconstruction from the steel-tube CR images. Parameters related to radiation and measuring structure, such as intensities, absorption in material and geometric specifications linked with the collimating components, are calculated and identified in order to construct the renoval images for twofold regions of circle-type steel tubes. A correction procedure for region recovery most similar to the true tube is designed. The application of this approach on CR images is shown and reconstructed results are discussed.

• Journal of Radiation Protection and Research
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• v.27 no.3
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• pp.171-179
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• 2002

The three-dimensional long-range dispersion model has been developed to understand the characteristics of the transport and diffusion of radioactive materials released into atmosphere. The model is designed to compute air concentration and ground deposition at distances up to some thousands of kilometers from the source point in horizontal direction. The vertical turbulent motion is considered separately within the mixing layer and above the mixing layer. The test simulation was performed In the area of Northeast Asia. The release point was assumed in the east part of China. The calculated concentration distributions art mainly advected toward the southeast part of release point by the wind fields. The developed model will be used to estimate the radiological consequences against a nuclear accident. The model will be supplemented by the comparative study using the data of the long-range field experiments.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.368-372
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• 2016

Background: Methodologies for a series of radiological consequence assessments show a distinctive difference according to the design principles of the original nuclear suppliers and their technical standards to be imposed. This is due to the uncertainties of the accidental source term, radionuclide behavior in the environment, and subsequent radiological dose. Both types of PWR and PHWR are operated in Korea. However, technical standards for evaluating atmospheric dispersion have been enacted based on the U.S. NRC's positions regardless of the reactor types. For this reason, it might cause a controversy between the licensor and licensee of a nuclear power plant. Materials and Methods: It was modelled under the framework of the NRC Regulatory Guide 1.145 for light-water reactors, reflecting the features of heavy-water reactors as specified in the Canadian National Standard and the modelling features in MACCS2, such as atmospheric diffusion coefficient, ground deposition, surface roughness, radioactive plume depletion, and exposure from ground deposition. Results and Discussion: An integrated accident consequence assessment code, ACCESS (Accident Consequence Assessment Code for Evaluating Site Suitability), was developed by taking into account the unique regulatory positions for reactor types under the framework of the current Korean technical standards. Field tracer experiments and hand calculations have been carried out for validation and verification of the models. Conclusion: The modelling approaches of ACCESS and its features are introduced, and its applicative results for a hypothetical accidental scenario are comprehensively discussed. In an applicative study, the predicted results by the light-water reactor assessment model were higher than those by other models in terms of total doses.

• Nuclear Engineering and Technology
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• v.2 no.2
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• pp.97-106
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• 1970

Radiation accidents which occured in the A.E.R.I. during last ten years are described (table 1). It seemed to the authors that some of these accidents were considered to be hazardous to man body and associated installations. This report deals with the following four major accidents involving body contamination incidents that our health physicists have been experienced. 1. Over-exposures (up to 130 rem) to the total body due to the mismanipulation in the Cobalt-60 gamma irradiation facility. 2. Floor surface contamination (up to 13 mrad/hr) and its spread out due to the mishandling of radioiodine contained in the bottle. 3. Body surface contamination and 0.36 uCi radioactivity accumulated in the thyroid gland of a worker due to the inhalation of gaseous iodine-131. 4. A void capsule due to the leakage out of the radium therapeutic source (3mg\ulcorner) These accidents were treated by definitely prompt action to protect the workers and associated installations from any radiation hazards and every possible efforts were made to confine the spread of radioactive contamination as small area as possible by means of elaborate decontamination work and monitoring.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2361-2369
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• 2020

Compton cameras using large scintillators have been developed for high imaging sensitivity. These scintillator-based Compton cameras, however, mainly due to relatively low energy resolution, suffer from undesired background-radiation signals, especially when radioactive materials' activity is very low or their location is far from the Compton camera. To alleviate this problem for a large-size Compton camera, in the present study, a hybrid-type shielding system was designed that combines an active shield with a veto detector and a passive shield that surrounds the active shield. Then, the performance of the hybrid shielding system was predicted, by Monte Carlo radiation transport simulation using Geant4, in terms of minimum detectable activity (MDA), signal-to-noise ratio (SNR), and image resolution. Our simulation results show that, for the most cases, the hybrid shielding system significantly improves the performance of the large-size Compton camera. For the cases investigated in the present study, the use of the shielding system decreased the MDA by about 1.4, 1.6, and 1.3 times, increased the SNR by 1.2-1.9, 1.1-1.7, and 1.3-2.1 times, and improved the image resolution (i.e., reduced the FWHM) by 7-8, 1-6, and 3-5% for ¹³⁷Cs, ⁶⁰Co, and ¹³¹I point source located at 1-5 m from the imaging system, respectively.

• Tunnel and Underground Space
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• v.21 no.5
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• pp.349-358
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• 2011

This paper introduces a new P wave arrival time determination algorithm of acoustic emission (AE) suitable to identify P waves with low signal-to-noise ratio generated in rock masses around the high-level radioactive waste disposal repositories. The algorithms adopted for this paper were amplitude threshold picker, Akaike Information Criterion (AIC), two step AIC, and Hinkley criterion. The elastic waves were generated by Pencil Lead Break test on a granite sample, then mixed with white noise to make it difficult to distinguish P wave artificially. The results obtained from amplitude threshold picker, AIC, and Hinkley criterion produced relatively large error due to the low signal-to-noise ratio. On the other hand, two step AIC algorithm provided the correct results regardless of white noise so that the accuracy of source localization was more improved and could be satisfied with the error range.