• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.671-675
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• 2003

In accordance with the KRR-1 & 2 decommissioning project, the decontamination and dismantling activities of the KRR-2 auxiliary facilities, radioisotope production facilities, were completed from Aug 2001 to Dec 2002. The auxiliary facilities were composed of the concrete hot-cell, lead hot-cells and several laboratories for the radioisotope production. The dismantling objects are home hoods, experimental desks, sinks, and contaminated inner facilities. For the purpose of the safe decommissioning activity, the method statements and working procedures were set up. The manpower of the total 20,933 man-hour was required and several dismantling equipments were also. The maximum surface contamination is: 9.24 Bq/$\textrm{cm}^2$ in removable contamination and 350,000 cpm in fixed contamination. The total amount of 62.146 Ton was raised as dismantled waste with kinds of the concretes, wood, steels, etc. The collective dose was evaluated as 0.33 mam-mSv during this period.

• Journal of Radiation Protection and Research
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• v.38 no.2
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• pp.52-59
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• 2013

The internal contamination screening method using gross beta measurement was performed for radioisotope workers. 24 h and spot urine samples from workers of medical isotope production facilities were collected and measured. Most of the results were similar with the background level of gross beta activity except for a specific worker. Gross beta activity was slightly increased in several hours after finishing work. And the environmental factor of production facilities causing internal contamination were estimated based on screening results. The additional detailed internal dose assessment must be followed after the screening for protection of workers. Moreover, a procedure was established to apply a simple internal contamination assessment for radiation workers.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.3996-4001
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• 2023

The High-flux Advanced Neutron Application Reactor (HANARO) produces radioisotopes (RIs) (¹³¹I, ¹⁹²Ir, etc.) through neutron irradiation on various RI production targets. Among them, ¹⁷⁷Lu and ¹⁶⁶Ho are particularly promising owing to their theranostic characteristics that facilitate simultaneous diagnosis and treatment. Prior to neutron irradiation, evaluating the nuclear heating of the RI production target is essential for ensuring the thermal-hydraulic safety of HANARO. In this study, the feasibility of producing ¹⁷⁷Lu and ¹⁶⁶Ho using irradiation holes of HANARO was investigated in terms of thermal-hydraulic safety. The nuclear heating rates of the RI production target by prompt and delayed radiation were calculated using MCNP6. The calculated nuclear heating rates were used as an input parameter in COMSOL Multiphysics to obtain the temperature distribution in an irradiation hole. The degree of temperature increase of the ¹⁷⁷Lu and ¹⁶⁶Ho production targets satisfied the safety criteria of HANARO. The nuclear heating rates and temperature distribution obtained through the in silico study are expected to provide valuable insight into the production of ¹⁷⁷Lu and ¹⁶⁶Ho using HANARO.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.113-122
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• 1990

After four years of planning, equipment acquisition, facility construction and beam testing, the KCCH cyclotron facility was put into operation in November1986. Now the KCCH cyclotron(MC-50) has been used for four years in neutron therapy and radioisotope production. Up to December 1989, 179(1852 sessions) patient have undergone neutron therapy. Radioisotope production for nuclear medicine use was started from March 1989 after extensive work to overcome target transport, target melting, beam diagnostic and chemical processing problems. This status report introduces the cyclotron facility, and the experiences of neutron therapy and isotope production with the MC-50 cyclotron. Besides, the operation results and the general troubles of the MC-50 during 1989 are summarized. Total operation time was 1252.5 hours. Four hundred hours were used for neutron therapy of 599 treatment sessions and 832.5 hours for radioisotope production. Total amount of produced raioisotope was 1695 mCi(Ga-67 : 1478mCi, Tl-201 : 107 mCi, I-123 : 25mCi, In-111 : 85mCi). Twenty hours were used for scheduled beam testing. In 1989, 882% of the planned operation were performed on schedule and this rats is improved remarkably compared to 71.0% in 1988.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.387.2-387.2
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• 2003

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1355-1363
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• 2018

Many radionuclides exist in normal environment and artificial radionuclides also can be detected. The radionuclides ($^{131}I$) are widely used for labeling compounds and radiation therapy. In Korea, the radionuclide ($^{131}I$) is produced at the Radioisotope Production Facility (RIPF) at the Korea Atomic Energy Research Institute in Daejeon. The residents around the RIPF assume that $^{131}I$ detected in environmental samples is produced from RIPF. To ensure the safety of the residents, the radioactive concentration of $^{131}I$ near the RIPF was investigated by monitoring environmental samples along the Gap River. The selected geographical places are near the nuclear installation, another possible location for $^{131}I$ detection, and downstream of the Gap River. The first selected places are the "front gate of KAERI", and the "Donghwa bridge". The second selected place is the sewage treatment plant. Therefore, the Wonchon bridge is selected for the upstream of the plant and the sewage treatment plant is selected for the downstream of the plant. The last selected places are the downstream where the two paths converged, which is Yongshin bridge (in front of the cogeneration plant). In these places, environmental samples, including sediment, fish, surface water, and aquatic plants, were collected. In this study, the radioactive iodine ($^{131}I$) detection along the Gap River will be investigated.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.20 no.5
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• pp.467-470
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• 2007

Application field of neutron beam is very broad including industry, medicine and science. But the research and development and use of neutron beam is restricted within in narrow limits in this country, because neutron beam facility is insufficient - a big research facility of nuclear reactor(HANARO) and some small industrial facilities which use radioisotope neutron source are available. This paper compare and investigate the results of experiment and numerical analysis of the discharge in the spherically convergent beam fusion device which were expected as a portable neutron source. The spherically convergent beam fusion device will offer stability in neutron production, possibility of movement for convenience, low construction cost and higher neutron flux than radioisotope neutron source. The star mode discharge which efficiently generate neutron, were observed at both results.

• Journal of Radiation Industry
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• v.8 no.3
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• pp.141-146
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• 2014

Nickel (Ni) electroplating was implemented by using a metal Ni powder in order to establish a $^{63}Ni$ plating condition on the PN junction semiconductor needed for production of beta-voltaic battery. PN junction semiconductors with a Ni seed layer of 500 and $1000{\AA}$ were coated with Ni at current density from 10 to $50mA\;cm^{-2}$. The surface roughness and average grain size of Ni deposits were investigated by XRD and SEM techniques. The roughness of Ni deposit was increased as the current density was increased, and decreased as the thickness of Ni seed layer was increased. The results showed that the optimum surface shape was obtained at a current density of $10mA\;cm^{-2}$ in seed layer with thickness of $500{\AA}$, $20mA\;cm^{-2}$ of $1000{\AA}$. Also, pure Ni deposit was well coated on a PN junction semiconductor without any oxide forms. Using the line width of (111) in XRD peak, the average grain size of the Ni thick firm was measured. The results showed that the average grain size was increased as the thickness of seed layer was increased.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.979-986
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• 2013

Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99) production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 $g-U/cm^3$ were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compounds by a chemical reaction of the uranium particles and aluminum matrix. Thus, these target plates can be treated with the same alkaline dissolution process that is used for conventional $UAl_x$ dispersion targets, while increasing the uranium density in the target plates.