• The Korean Journal of Nuclear Medicine
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• v.34 no.1
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• pp.62-73
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• 2000

Purpose: Esophageal cancer patients have a difficulty in the intake of meals through the blocked esophageal lumen, which is caused by an ingrowth of cancer cells and largely influences on the prognosis. It is reported that esophageal cancer has a very low survival rate due to the lack of nourishment and immunity as the result of this. In this study a new radioactive stent, which prevents tumor ingrowth and restenosis by additional radiation treatment, has been developed. Materials and Methods: Using ${\ulcorner}HANARO{\lrcorner}$ research reactor, the radioactive stent assembly ($^{166}Ho$-SA) was prepared by covering the metallic stent with a radioactive sleeve by means of a post-irradiation and pre-irradiation methods. Results: Scanning electron microscopy and autoradiography exhibited that the distribution of $^{165/166}Ho\;(NO_3)$ compounds in polyurethane matrix was homogeneous. A geometrical model of the esophagus considering its structural properties, was developed for the computer simulation of energy deposition to the esophageal wall. The dose distributions of $^{166}Ho$-stent were calculated by means of the EGS4 code system. The sources are considered to be distributed uniformly on the surface in the form of a cylinder with a diameter of 20 mm and length of 40 mm. As an animal experiment, when radioactive stent developed in this study was inserted into the esophagus of a Mongrel dog, tissue destruction and widening of the esophageal lumen were observed. Conclusion: We have developed a new radioactive stent comprising of a radioactive tubular sleeve covering the metallic stent, which emits homogeneous radiation. If it is inserted into the blocked or narrowed lumen, it can lead to local destruction of the tumor due to irradiation effect with dilatation resulting from self-expansion of the metallic property. Accordingly, it is expected that restenosis esophageal lumen by the continuous ingrowth and infiltration of cancer after insertion of our radioactive stent will be decreased remarkably.

• Progress in Medical Physics
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• v.16 no.1
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• pp.16-23
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• 2005

The objective of this study was to construct multi-element ultrasound applicators for the treatment of gynecologic cancer with high dose rate brachytherapy. For the treatment of uterus, piezo-ceramic crystal transducer (PZT -5A) with outer diameter of 4 mm, wall thickness of 1.3 mm, and length of 24.5 mm was selected. For the treatment of cervix or vagina, it should be possible to insert the applicator into the vagina. Thus, a cylindrical PZT -8 material with outer diameter of 24.5 mm, wall thickness of 1.3 mm, and length of 15.2 mm was selected. The operating frequencies determined by vector impedance measurement were 3.2 MHz for the PZT 5A cylinder (OD=4 mm) and 1.7 MHz for the PZT -8 cylinder (OD: 24.5 mm). The ratios of generated acoustic output power to applied electric power were 33% and 61% for the tandem type crystal and the cylinder type crystal, respectively. The radiated acoustic pressure fields from both transducers were calculated using a Matlab code and measured in water using hydrophone. There was good agreement between measured and calculated acoustic pressure field distribution. For a tandem type transducer, the calculated acoustic pressure field decreased from 0.023 MPa at 10 mm to 0.010 Mpa at 30 mm, the reduction of 57%. For the cylinder type transducer which will be used for the treatment of vagina showed 78% reduction at 15 mm and 66% at 25 mm as compared to values at 5 mm from the surface. Based on the characteristics of the transducers, this study demonstrated the possibility of using the crystals as a heating source. Finally, a 3-element and 4-element prototype applicators were constructed. The 3-element applicator is 75 mm long and 4 mm thick and will be used for the treatment of uterus. The 4-element applicator is 61 mm long and 24.5 mm thick and will be used for the treatment of vagina. Using these applicators, it is possible to generate enough power to increase temperature to therapeutic level.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.305-315
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• 2020

For geological disposal of radioactive wastes, a method was proposed to evaluate the radionuclide transport in the biosphere by calculating the elapsed time of nuclide migration. The radionuclides were supposed to be introduced from a natural barrier and reached a large surface water body following a groundwater flow in a shallow subsurface. The biosphere was defined as a shallow subsurface environment that included aquifers on a host rock. Using the proposed method, a calculation algorithm was established, and a computer code that implemented the algorithm was developed. The developed code was verified by comparing the simulation results of the simple cases with the results of the analytical solution and a public program, which has been widely used to evaluate the radiation dose using the radionuclide transport near the surface. A case study was constructed using the previous research for radionuclide transport from the hypothetical geological disposal repository. In the case study, the code calculated the mass discharge rate of radionuclide to a stream in the biosphere. Because the previous research only demonstrated the transport of radionuclides from the hypothetical repository to the host rock, the developed code in the present study could help identify the total transport of radionuclide along the complete pathway.

• Journal of radiological science and technology
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• v.42 no.6
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• pp.447-454
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• 2019

In order to improve and supplement the shielding method for electron beam treatment, we designed a patient-specific shielding method using a 3D printer, and evaluated the usefulness by comparing and analyzing the distribution of electron beam doses to adjacent organs. In order to treat 5 cm sized superficial tumors around the lens, a CT Simulator was used to scan the Alderson Rando phantom and the DICOM file was converted into an STL file. The converted STL file was used to design a patient-specific shield and mold that matched the body surface contour of the treatment site. The thickness of the shield was 1 cm and 1.5 cm, and the mold was printed using a 3D printer, and the patient customized shielding block (PCSB) was fabricated with a cerrobend alloy with a thickness of 1 cm and 1.5 cm. The dosimetry was performed by attaching an EBT3 film on the surface of the Alderson Rando phantom eyelid and measuring the dose of 6, 9, and 12 MeV electron beams on the film using four shielding methods. Shielding rates were 83.89%, 87.14%, 87.39% at 6, 9, and 12 MeV without shielding, 1 cm (92.04%, 87.48%, 86.49%), 1.5 cm (91.13%, 91.88% with PSCB), 92.66%) The shielding rate was measured as 1 cm (90.7%, 92.23%, 88.08%) and 1.5 cm (88.31%, 90.66%, 91.81%) when the shielding block and the patient-specific shield were used together. PCSB fabrication improves shielding efficiency over conventional shielding methods. Therefore, PSCB may be useful for clinical application.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.