• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.215-216
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• 2010

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.135-143
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• 2017

The domestic unique research reactor, HANARO (Hi-flux Advanced Neutron Application ReactOr), has been constructed with the open-pool, the core is submerged in, for the multi-purpose neutron application. The reactor has a primary cooling system to remove the fission heat from the core and its connected fluidic systems. Since the works are required at the reactor pool top as a characteristic of the research reactor, the radiation shall be minimized with the operation of the hot water layer system to avoid unnecessary radiation exposure on the workers during work at the pool top. Moreover, the pool water management system is connected to the reactor pool to maintain the pool temperature below $50^{\circ}C$ to minimize the uprising radioactive gas or impurity from the colder pool bottom. For the efficient flow rate of the PWMS, the thermal capacity of heat exchanger is selected with 260 kW in the normal operation condition. In this paper, the modeling is formulated to figure out whether or not each pool temperature maintains below the temperature limit and the calculation results show that the designed PWMS heat exchanger has enough capacity with the design margin regardless of the reactor operation mode.

• Journal of Radiation Industry
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• v.9 no.2
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• pp.63-68
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• 2015

Lyophilized DOTMP kits were prepared using DOTMP, ammonium acetate, and ascorbic acid. The $^{68}Ga$-DOTMP was prepared by incubating the kit dissolved in 0.5 ml of concentrated $^{68}Ga$ using NaCl method and 0.5 ml of DDW, at $100^{\circ}C$ for 7 min. The labeling yield was evaluated by two solvent systems of TLC. 1 MBq of concentrated $^{68}Ga$ was labeled with $0.8{\mu}g$ of DOTMP by high radiolabeling yield (>98%), which was determined by two TLC methods. The composition of the prepared freeze-dried vial is $400{\mu}g$ of DOTMP, 19.27 mg of ammonium acetate and 17.62 mg of ascorbic acid. ~555 MBq of $^{68}Ga$-DOTMP was prepared with excellent radiochemical purity (>98%) and it was stable for 4 hr at room temperature. In conclusion, Freeze-dried DOTMP kits for the convenient preparation of $^{68}Ga$-DOTMP have been developed. Availability of this kit is expected to stimulate the widespread use of $^{68}Ga$-DOTMP in the fields of nuclear medicine.

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.365-374
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• 2001

In order to examine the in-reactor performance of very-high-density dispersion fuels for high flux performance research reactors, U-l0wt.%Mo microplates containing centrifugally atomized powder were irradiated at low temperature. The U-l0wt.%Mo dispersion fuels show stable in- reactor irradiation behaviors even at high burn-up, similar to U$_3$Si$_2$ dispersion fuels. The atomized U-l0wt.%Mo fuel particles have a fine and a relatively uniform fission gas bubble size distribution. Moreover, only one of third of the area of the atomized fuel cross-sections at 70a1.% burn-up shows fission gas bubble-free zones, This appears to be the result of segregation into high Mo and low Mo.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4226-4230
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• 2022

The Korea Atomic Energy Research Institute (KAERI) has designed a Hybrid-Low Power Research Reactor (H-LPRR) which can be used for critical assembly and conventional research reactor as well. It is an open tank-in-pool type research reactor (Thermal Power: 50 kW_th) of which the most important applications are Neutron Activation Analysis (NAA), Radioisotope (RI) production, education and training. There are eight irradiation holes on the edge of the reactor core: IR (6 holes for RI production) and NA (2 holes for NAA) holes. In order to quantify the elemental concentration in target samples through the Instrumental Neutron Activation Analysis (INAA), it is necessary to measure neutron spectrum parameters such as thermal neutron flux, the deviation from the ideal 1/E epithermal neutron flux distribution (α), and the thermal-to-epithermal neutron flux ratio (f) for the irradiation holes. In this study, the MCNP6.1 code and FORTRAN 90 language are applied to determine the parameters for the two irradiation holes (NA-SW and NA-NW) in H-LPRR, and in particular its α and f parameters are compared to values of other research reactors. The results confirmed that the neutron irradiation holes in H-LPRR are designed to be sufficiently applied to neutron activation analysis, and its performance is comparable to that of foreign research reactors including the TRIGA MARK II.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.915-921
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• 2019

Radiation damage due to neutrons has been calculated in Ni-based alloys in Wolsong CANDU reactor environments. Two damage parameters are considered: displacement damage, and transmutation gas production. We used the SPECTER and SRIM computer codes in quantifying radiation damage. In addition, damage caused by Ni two-step reactions was considered. Estimations were made for the annulus spacers in a CANDU reactor that are located axially along a fuel channel and made of Inconel X-750. The calculation results indicate that the transmutation gas production from the Ni two-step reactions is predominant as the effective full power year increases. The displacement damage due to recoil atoms produced from Ni two-step reactions accounts for over 30% out of the total displacement damage.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.385-391
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• 2006

The estimated decommissioning cost of nuclear research reactor is calculated by applying a unit cost factor-based engineering cost calculation method on which classification of decommissioning works fitted with the features and specifications of decommissioning objects and establishment of composition factors are based. Decommissioning cost of nuclear research reactor is composed of labor cost, equipment and materials cost. Labor cost of decommissioning costs in decommissioning works are calculated on the basis of working time consumed in decommissioning objects. In this paper, the unit cost factors and work difficulty factors which are needed to calculate the labor cost in estimating decommissioning cost of nuclear research reactor are derived and figured out.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.532-533
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• 2005

• Transactions on Electrical and Electronic Materials
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• v.4 no.4
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• pp.30-35
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• 2003

In this paper, the $CF_4$ decomposition efficiency is investigated for three simulated plasma reactors that are needle plate reactor, metal particle reactor, and dielectric barrier reactor with hole (DBH). The$CF_4$ decomposition efficiency by DBH is much better than that by needle plate reactor or metal particle reactor. When applied voltage is increased up to the critical voltage for spark formation in the all reactors, the $CF_4$ decomposition efficiency is increased. The $CF_4$ decomposition efficiency in needle plate reactor and metal particle reactor is about $12\%$ and $22\%$ respectively at applied voltage of 23 kV (consumption power: 110 W) and $CF_4$ concentration of 500 ppm, however, the $CF_4$ decomposition efficiency is more than $95\%$ in case of DBH. DBH should be much better than two reactors investigated for $CF_4$ decomposition.