• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.304-317
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• 2016

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.14-18
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• 2020

Korea Atomic Energy Research Institute (KAERI) has carried out various neutronics experiments in the commissioning stage of the Jordan Research and Training Reactor (JRTR), and this paper introduces the results of first criticality prediction and experiment for the JRTR. The Monte Carlo Code for Advanced Reactor Design and analysis (McCARD) with the ENDF/B-VII.0 nuclear library was used for prediction calculations in the process of the first criticality approach, which was performed to provide reference for the first criticality experiment. In the experiment, fuel loading was carried out by measuring the inverse multiplication factor (1/M) to predict the number of fuel assemblies at the first criticality, and the first critical was reached on April 25, 2016. Comparing the first criticality prediction and experiment, the calculated and measured CAR (Control Absorber Rod) heights for the first criticality were 575 mm and 570.5 mm, respectively, that is, the difference between the two results was approximately 5 mm. From this result, it was confirmed that JRTR manufacturing and various experiments had successfully progressed as designed.

• 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.55 no.5
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• pp.1814-1820
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• 2023

Safety classification of systems, structures, and components (SSC) is an essential activity for nuclear reactor design and operation. The current regulatory trend is to require risk-informed safety classification that considers first, the severity, but also the frequency of SSC failures. While safety classification for nuclear power plants is covered in many regulatory and scientific publications, research reactors received less attention. Research reactors are typically of lower power but, at the same time, are less standardized i.e., have more variability in the design, operational modes, and operating conditions. This makes them more challenging when considering safety classification. This work presents the Integrated Risk-Informed Safety Classification (IRISC) procedure which is a novel extension of the IAEA recommended process with dedicated probabilistic treatment of research reactor designs. The article provides the details of probabilistic analysis performed within safety classification process to a degree that is often missing in most literature on the topic. The article presents insight from the implementation of the procedure in the safety classification for the MARIA Research Reactor operated by the National Center for Nuclear Research in Poland.

• 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

• 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.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.216-221
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• 2016

Background: Very fine radiation-induced aerosol particles are produced in intense radiation fields, such as high-intensity accelerator rooms and containment vessels such as those in the Fukushima Daiichi nuclear power plant (FDNPP). Size measurement of the aerosol particles is very important for understanding the behavior of radioactive aerosols released in the FDNPP accident and radiation safety in high-energy accelerators. Materials and Methods: A combined technique using wire screens and imaging plates was developed for size measurement of fine radioactive aerosol particles smaller than 100 nm in diameter. This technique was applied to the radiation field of a proton accelerator room, in which radioactive atoms produced in air during machine operation are incorporated into radiation-induced aerosol particles. The size of $^{11}C$-bearing aerosol particles was analyzed using the wire screen technique in distinction from other positron emitters in combination with a radioactive decay analysis. Results and Discussion: The size distribution for $^{11}C$-bearing aerosol particles was found to be ca. $70{\mu}m$ in geometric mean diameter. The size was similar to that for $^7Be$-bearing particles obtained by a Ge detector measurement, and was slightly larger than the number-based size distribution measured with a scanning mobility particle sizer. Conclusion: The particle size measuring method using wire screens and imaging plates was successfully applied to the fine aerosol particles produced in an intense radiation field of a proton accelerator. This technique is applicable to size measurement of radioactive aerosol particles produced in the intense radiation fields of radiation facilities.

• 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.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.191-202
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• 2013

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Proceedings of the Korean Vacuum Society Conference
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• 2015.08a
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• pp.104-104
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• 2015

Due to the demand of the cold neutron flux in the neutron science and beam utilization technology, the cold neutron source (CNS) has been constructed and operating in the nuclear research reactor all over the world. The majority of the heat load removal scheme in the CNS is two-phase thermosiphon using the liquid hydrogen as a moderator. The CNS moderates thermal neutrons through a cryogenic moderator, liquid hydrogen, into cold neutrons with the generation of the nuclear heat load. The liquid hydrogen in a moderator cell is evaporated for the removal of the generated heat load from the neutron moderation and flows upward into a heat exchanger, where the hydrogen gas is liquefied by the cryogenic helium gas supplied from a helium refrigeration system. The liquefied hydrogen flows down to the moderator cell. To keep the required liquid hydrogen stable in the moderator cell, the CNS consists of an in-pool assembly (IPA) connected with the hydrogen system to handle the required hydrogen gas, the vacuum system to create the thermal insulation, and the helium refrigeration system to provide the cooling capacity. If one of systems is running out of order, the operating research reactor shall be tripped because the integrity of the CNS-IPA is not secured under the full power operation of the reactor. To prevent unscheduled reactor shutdown during a long time because the research reactor has been operating with the multi-purposes, the introduction of the standby cooling system (STS) can be a solution. In this presentation, the design considerations are considered how to design the STS satisfied with the following objectives: (a) to keep the moderator cell less than 350 K during the full power operation of the reactor under loss of the vacuum, loss of the cooling power, loss of common electrical power, or loss of instrument air cases; (b) to circulate smoothly helium gas in the STS circulation loop; (c) to re-start-up the reactor within 1 hour after its trip to avoid the Xenon build-up because more than certain concentration of Xenon makes that the reactor cannot start-up again; (d) to minimize the possibility of the hydrogen-oxygen reaction in the hydrogen boundary.