• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1439-1447
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• 2023

The fuel cycle characteristics of the IVG.1M reactor were studied within the framework of the research reactor conversion program to modernize the IVG.1M reactor. Optimum use of the nuclear fuel and reactor was achieved through routine methods which included partial fuel reloading combined with scheduled maintenance operations. Since, the additional problem in planning the fuel cycle of the IVG.1M reactor was the poisoning of the beryllium parts of the core, reflector, and control system. An assessment of the residual power and composition of spent fuel is necessary for the selection and justification of the technology for its subsequent management. Computational studies were performed using the MCNP6.1 program and the neutronics model of the IVG.1M reactor. The proposed scheme of annual partial fuel reloading allows for maintaining a high reactor reactivity margin, stabilizing it within 2-4 β_eff for 20 years, and achieving a burnup of 9.9-10.8 MW × day/kg U in the steady state mode of fuel reloading. Spent fuel immediately after unloading from the reactor can be placed in a transport packaging cask for shipping or safely stored in dry storage at the research reactor site.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.764-769
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• 2022

The United Arab Emirates is currently building and operating four units of the APR-1400 developed by a South Korean vendor, Korea Electric Power Corporation (KEPCO). This paper attempts to perform APR-1400 reactor core analysis by using the well-known two-step method. The two-step method was applied to the APR-1400 first cycle using the open-source nodal diffusion code, KOMODO. In this study, the group constants were generated using CASMO-4 fuel transport lattice code. The simulation was performed in Hot Zero Power (HZP) at steady-state and transient conditions. Some typical parameters necessary for the Nuclear Design Report (NDR) were evaluated in this paper, such as effective neutron multiplication factor, control rod worth, and critical boron concentration for steady-state analysis. Other parameters such as reactivity insertion, power, and fuel temperature changes during the Reactivity Insertion Accident (RIA) simulation were evaluated as well. The results from KOMODO were verified using PARCS and SIMULATE-3 nodal core simulators. It was found that KOMODO gives an excellent agreement.

• 한국전산유체공학회:학술대회논문집
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• 2004.03a
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• pp.208-213
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• 2004

The HANARO, multi-purpose research reactor, 30 MWth open-tank-in- pool type, is under normal operation since it reached the initial critical in February 1995. The HANARO is planning to produce a fission moly-99 of radio isotopes, a mother nuclide of Tc-99m, a medical isotope and is under developing a target handling tool for loading and unloading it in a circular flow tube (OR-5). A guide tube is extended from the reactor core to the top of the reactor chimney for easily un/loading a target under the reactor normal operation. But active coolant through the core can be quickly raised up to the top of the chimney through the guide tube by jet flow. This paper is described an analytical analysis to calculate the hole size of a orifice inserted in the circular irradiation hole and to study the flow characteristics through the guide tube under reactor normal operation and loading the target. As results, the results show that the hole size of orifice was 31 mm of the inner diameter to suppress the guide tube jet flow and the coolant safely cooled the target of fission moly after inserting the orifice to the flow tube.

• 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.54 no.10
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• pp.3803-3810
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• 2022

OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. The Weight Window Mesh (WWM) function and an automatic Global Variance Reduction (GVR) method was recently developed and implemented in a developmental branch of OpenMC. This WWM function and GVR method broaden OpenMC's usage in general purposes deep penetration shielding calculations. However, the Local Variance Reduction (LVR) method, which suits the source-detector problem, is still missing in OpenMC. In this work, the Weight Window Generator (WWG) function has been developed and benchmarked for the same branch. This WWG function allows OpenMC to generate the WWM for the source-detector problem on its own. Single-material cases with varying shielding and sources were used to benchmark the WWG function and investigate how to set up the particle histories utilized in WWG-run and WWM-run. Results show that there is a maximum improvement of WWM generated by WWG. Based on the above results, instructions on determining the particle histories utilized in WWG-run and WWM-run for optimal computation efficiency are given and tested with a few multi-material cases. These benchmarks demonstrate the ability of the OpenMC WWG function and the above instructions for the source-detector problem. This developmental branch will be released and merged into the main distribution in the future.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.469-476
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• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• The KSFM Journal of Fluid Machinery
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• v.5 no.1 s.14
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• pp.49-54
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• 2002

HANARO, 30 MW of research reactor, was installed at the depth of 13m in an open pool. The $90\%$ of primary coolant was designed to pass through the core and to remove the reaction heat of the cote. The rest, $10\%$, of the primary coolant was designed to bypass the core. And the reactor coolant through and bypass the core was inhaled at the top of chimney by the coolant pump to prevent the radiated gas from being lifted to the top of reactor pool. But, the part of core bypass coolant was not inhaled by the reactor coolant pump and reached at the top of reactor pool by natural convection, and increased the radiation lovel on the top of reactor pool. To reduce the radiation level by protecting the natural convection of the core bypass flow, the hot water layer (HWL, hereinafter) was installed with the depth of 1.2 m from the top of reactor pool. As the HWL was normally operated, the radiation level was reduced to five percent ($5\%$) in comparing with that before the installation of the HWL. When HANARO was operated at a higher temperature than the normal temperature of the HWL by operating the standby heater, it was found that the radiation level was more reduced than that before operation. To verify the reason, the heat loss of the HWL was calculated by Visual Basic Program. It was confirmed through the results that the larger the temperature difference between the HWL and reactor hall was, the more the evaporation loss increased. And it was verified that the radiation level above was reduced mote safely by increasing the capacity of heater.

• 유체기계공업학회:학술대회논문집
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• 2001.11a
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• pp.221-226
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• 2001

HANARO, 30MW of research reactor, was installed at the depth of 13m of open pool, The $90\%$ of primary coolant was designed to pass through the core and to remove the reaction heat of the core. The rest $10\%$, of the primary coolant was designed to bypass the core. And the reactor coolant through and bypass the core was inhaled at the top of chimney by the coolant pump to protect that the radiated gas was lifted to the top of reactor pool. But, the part of core bypass coolant was not inhaled by the reactor coolant pump and reached at the top of reactor pool by natural convection and increased the radiation level on the top of reactor pool. To reduce the radiation level by protecting the natural convection of the core bypass flow, the hot water layer (HWL, hereinafter) was installed with the depth of 1.2m from the top of reactor pool. As the HWL was normally operated, the radiation level was reduced to five percent ($5\%$) in comparing with that before the installation of the HWL. When HANARO was operated with higher temperature than the normal temperature of the HWL by operating the standby heater, it was found that the radiation level was more reduced than that before operation. To verify the reason, the heat loss of the HWL was calculated. It was confirmed through the results that the larger the temperature difference between the HWL and reactor hall was, the more the evaporation loss was increased. And it was verified that the radiation level above was reduced more safely by increasing the capacity of heater.

• Journal of the Korean Society of Safety
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• v.11 no.3
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• pp.112-118
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• 1996

When fossil fuels are burned they produce CO gas because of incomplete combustion. If the CO gas reacts with the hemoglobin in the red blood cells, it may result in death or sequelae. Generally, the CO gas is eliminated in the form of the $$$CO_2$ gas by the oxidation reaction over the platinum catalyst. In this study, the electrochemical CO removal was investgated by using the solid electrolyte cell reactor, the type of which was represented as reactants$/Pt/Y_2O_3-ZrO_2/Pt/Air$. If the overpotential was applied to the platinum working electrode, the conversion could be changed with the overpotential applied. It was found that the oxidation rate could be increased 2.8 times higher than that of the normal condition, i. e. under open circuit conditions when $P_{co}/P_{O_2}$ was 0.5 and overpotential was 0.9V. From these results, it is concluded that the reactor used in this study is more efficient than conventional catalytic reactors.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.19 no.1
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• pp.169-174
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• 2005

The conventional harmonic filters to reduce zero harmonic current from neutral reactors and Zig-Zag connection, have several disadvantages of the decreased reduction rate of harmonics under a light load, because they have the load factor-dependent reduction rate of harmonics, and the risk of potential breaking in the neutral line by heated neutral reactor. Based Zig-Zag connection and Open-Delta mode, this new harmonic eliminating device (HANOS) adopts the combination of Zig-Zag connection and Open-Delta mode-the latter is additionally applied to the transformer's core block for connection to the neutral line. The results of this study demonstrated that the new device could eliminate safely zero harmonic current running in the neutral line without heating.