• Journal of Aerospace System Engineering
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• v.16 no.3
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• pp.73-83
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• 2022

The purpose of this paper was to identify a method to improve received noise in a specific frequency band, caused by RF radiation signals during aircraft operation. The communication equipment of the aircraft is critical to the performance and safety of the flight mission, as it is responsible for the function of internal and external communications. Noise-free, clean communication quality, as well as transmission and receiving functions have to be implemented. Thus, the cause analysis of the reception noise in a specific frequency band, was analyzed through trouble shooting. The receiving noise due to the RF radiation signal emitted from the subsystem unit inside the aircraft, was improved by shielding with a CAP with electroless nickel plating applied. Additionally, the measurement and verification results of the improvement method are also described.

• Journal of Radiation Protection and Research
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• v.38 no.4
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• pp.189-193
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• 2013

Annual dose on the containment building wall of the interim storage facility at normal condition was calculated to estimate the dose rate transition of the facility of PWR spent nuclear fuel. In this study, source term was generated by ORIGEN-ARP with 4.5 wt% initial enrichment, 45,000 MWd/MTU burnup and 10 years cooling time. Modeling of the storage facility and the containment building and radiation shielding evaluations were conducted by MCNP code depending on the distance between the wall and the facility in the building. In the case of the centralized storage system, the distance required for the annual dose rate limit from 10CFR72 was estimated to be 50 m.

• Journal of radiological science and technology
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• v.45 no.6
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• pp.523-529
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• 2022

Gamma spectroscopy analysis is widely used for radioactivity analysis, and various factors are required for radioactivity calculations. Among the factors, K3 for each sample significantly influences the results. The previous methods of correcting the self-absorption effect include a computational simulation method and a method that requires making a CRM(certified reference material) identical to the sample medium. However, the above methods have limitations when used in small institutions because they require specialized program utilization skills or high manufacturing costs and large facilities. The aim of this study is to develop a method that can be easily and rapidly applied to radioactivity analysis. After filling the beaker with water, we placed the radiation source in a uniform position and used the measured value as the benchmark. Next, a correction factor was derived based on the difference in the radiation source count of the benchmark and the identically measured sample. For the radiation source, Eu-152, which emits a broad range of energy within the measurement range of gamma rays, and Cs-134 and Cs-137, which are indicator nuclides in environmental radiation analysis, were used. The sample was selected within the density range of 0.26-2.11 g/cm³, and the correction factor was derived by calculating the count difference of each sample compared to the reference value of water. This study presents a faster and more convenient method than the existing research methods for determining the self-absorption effect correction, which has become increasingly necessary.

• Analytical Science and Technology
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• v.13 no.4
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• pp.474-483
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• 2000

An inductively coupled plasma atomic emission spectrometer/shielding system was specially designed and built for the analysis of radioactive materials. Both of an inductively coupled plasma source and a sample transfer system to be contacted with radioactive materials was installed in a stainless steel glove box. In terms of analytical capability and radiation safety, characteristic feature of the system was investigated. Its applicability to the determination of fission products and corrosion products in the radioactive materials such as spent fuel dissolver solution and the primary coolant of nuclear power reactors was evaluated. In the concentration range $0.01-0.1mgL^{-1}$, the relative standard deviation was found to be less than 5%.

• Journal of Radiation Protection and Research
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• v.40 no.3
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• pp.162-167
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• 2015

Although the perfomance indicators of the nuclear power plants in Korea show optimal, it requires detailed analysis and discussion centered on the radiation dose. As analysis methods, analysis on the radiation dose of nuclear power plants over the past five years was assessed by comparing the relevant radiation dose of radiation workers and per capita average annual radiation dose of the world's major nuclear power stations was also analyzed. The radiation workers over the annual radiation dose limit of 50 mSv were not. The contrast ratio of the radiation exposure according to the reactor type was the normal operation of PHWR was 6.2% higher than those of the PWR. This shows the radiation work of PHWR during normal driving operation is much more than those of PWR. According to the Performance Indicators of the World Association of Nuclear Operator, the annual radiation dose per unit in 2013 showed 527 man-mSv of Korea is the best country among the major nuclear power generating states, the world average was 725 man-mSv. The annual per capita radiation dose is about 80% less than 1 mSv of the public dose limit and also the average per capita dose showed a very low level as 0.82 mSv. Workers in related organizations showed 1.07 mSv, the non-destructive inspection agency workers showed 3.87 mSv. The remarkable results were due to radiation reduced program such as development of radiation shielding and radiation protection. In conclusion, the radiation exposured dose of nuclear power plants workers in Korea showed a trend which is ideally reduced. But more are expected to be difficul and the psychological insecurity against the operation of the nuclear power plants is existed to the residents near the nuclear power plants. So the radiation dose reduction policy and radiation dose follow up study of nuclear power plants will be continously excuted.

• Journal of Environmental Science International
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• v.3 no.1
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• pp.17-29
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• 1994

An one dimensional atmosphere-vegetation interaction model is developed to discuss of the effect of vegetation on heat flux in mesoscale planetary boundary layer. The canopy model was a coupled system of three balance equations of energy, moisture at ground surface and energy state of canopy with three independent variables of $T_f$(foliage temperature), $T_g$(ground temperature) and $q_g$(ground specific humidity). The model was verified by comparative study with OSUID(Oregon State University One Dimensional Model) proved in HYPEX-MOBHLY experiment. As the result, both vegetation and soil characteristics can be emphasized as an important factor iii the analysis of heat flux in the boundary layer. From the numerical experiments, following heat flux characteristics are clearly founded simulation. The larger shielding factor(vegetation) increase of $T_f$ while decrease $T_g$. because vegetation cut solar radiation to ground. Vegetation, the increase of roughness and resistance, increase of sensible heat flux in foliage while decrease the latent heat flux in the foliage.

• Progress in Medical Physics
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• v.12 no.2
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• pp.177-184
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• 2001

The goal of radiation treatment is to deliver a prescribed radiation dose to the target volume accurately while minimizing dose to normal tissues. Due to inaccurate placement of field and shielding block and patient's movement, there could be displacement errors between the planed and treatment regions. In order to verify the location of radiation treatment, we in this study developed the registration algorithm of the x-ray simulator images and portal images and quantified the inaccuracy in terms of shift, scale and rotation. The algorithm for registration of pairs of radiation fields consists of the alignment of pairs of radiation images by points matching and field displacement analysis by field boundary matching. In the first step, paired surface landmarks are matched to calculate the transformation parameters (scale, rotation and shift) using the corresponding line pairs which are created by connecting two landmarks of each image. In the next step, portal field boundary is extracted and then the two field boundaries are matched by the $\rho$-$\theta$ technique. Applying the phantom portal images, detection errors were calculated to be less than 2mm in translation, 1$^{\circ}$ in rotation and 1% in scale. In conclusion, we quantitatively analyzed the displacement error of x-ray simulator images and portal images. The present results could contribute to the study of the radiation treatment verification.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.73-80
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• 2014

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2999-3005
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• 2022

The new experimental setup has been built at the 11b channel of the IBR-2 research reactor at FLNP, JINR, to study the elemental composition of samples by registration of prompt gamma emission during thermal neutron capture. The setup consists of a curved mirror neutron guide and a radiation-resistant HPGe high-purity germanium detector. The detector is surrounded by lead shielding to suppress the natural background gamma level. The sample is placed in a vacuum channel and surrounded by a LiF shield to suppress the gamma background generated by scattered neutrons. This work presents characteristics of the experimental setup. An example of hydrogen concentration determining in a diamond powder made by detonation synthesis is given and on its basis, the sensitivity of the setup is calculated being ~4 ㎍.

• The Journal of Korean Institute of Electromagnetic Engineering and Science
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• v.20 no.9
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• pp.936-942
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• 2009

This paper presents the delivered power and reflection coefficient in metallic shielding enclosure with a sub-cavity, which are evaluated with the method of moments, sad describes a method for controlling the resonance characteristics of the metallic cavity by putting lossy dielectric material in the sub-cavity. In this paper we introduce carbon polystyrene-foam as lossy dielectric material and observe it's effects of reduction when the dimensions of the sub-cavity and permittivity of lossy dielectric material are changed. The results show that the reduction of the electromagnetic radiation can be achieved by controlling the amount of carbon in lossy dielectric material and the dimensions of the sub-cavity. The theoretical analysis is verified by the measured delivered power.