• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.647-653
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• 2019

A comprehensive study of photon interaction features has been made for some alloys containing Pd and Ag content to evaluate its possible use as alternative gamma radiations shielding material. The mass attenuation coefficient (${\mu}/{\rho}$) of the present alloys was measured at various photon energies between 81 keV - 1333 keV utilizing HPGe detector. The measured ${\mu}/{\rho}$ values were compared to those of theoretical and computational (MCNPX code) results. The results exhibited that the ${\mu}/{\rho}$ values of the studied alloys are in same line with results of WinXCOM software and MCNPX code results at all energies. Moreover, Pd75/Ag25 alloy sample has the maximum radiation protection efficiency (about 53% at 81 keV) and lowest half value layer, which shows that Pd75/Ag25 has superior gamma radiation shielding performance among the compared other alloys.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1272-1278
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• 2019

A lumped kinetic modeling platform is developed to investigate the coupled nuclear/thermo-fluid features of the closed natural circulation loop in a low power lead cooled fast reactor. This coolant material serves a reliable choice with noticeable thermo-physical safety characteristics in terms of natural convection. Boussienesq approximation is resorted to appropriately reduce the governing partial differential equations (PDEs) for the fluid flow into a set of ordinary differential equations (ODEs). As a main contributing step, the coolant circulation speed is accordingly correlated to the loop operational power and temperature levels. Further temporal analysis and control synthesis activities may thus be carried out within a more consistent state space framework. Nyquist stability criterion is thereafter employed to carry out a sensitivity analysis for the system stability at various power and heat sink temperature levels and results confirm a widely stable natural circulation loop.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1349-1354
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• 2018

Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES) is widely used for the determination of trace elements in geological samples in Iran. In this paper, we have calculated the detection limits of neutron activation analysis (NAA) for some of the common elements in such samples utilizing the ORIGEN and MCNP codes and verified the simulations using the experimental results of three soil standard reference materials, namely, G02.SRM, G18.SRM, and G28.SRM. The results show that while the detection limit of ICP-AES method is usually in the mg/kg range, it is represented to the ${\mu}g/kg$ range for most of the elements of interest using the NAA method, and the simulations can be verified in a tolerance range of 20%.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• The Journal of Korea Robotics Society
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• v.17 no.2
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• pp.178-182
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• 2022

Post-tension duct in nuclear reactor containment building is filled with grease to prevent steel strand from corroding. If grease leaks by break of duct, steel strand will corrode and cause problem in building safety. Therefore, grease leak should be checked preventatively. But currently used method is inefficient, since it has to remove grease and strand to check. And other methods for checking post-tension dust are not well-researched. In this paper, we develop screw-propelled robot that can move in grease and detect grease duct directly. Also, we make the test environment that is similar to real post-tension duct of containment building and test robot in that environment to verify that our robot is feasible in the post-tension duct. The robot can move forward and backward in grease duct by twin screw mechanism and has mount for sensors to detect grease leakage and strand corrosion.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1635-1643
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• 2022

STPA (System-Theoretic Process Analysis) is a widely used safety analysis technique to identify UCAs (Unsafe Control Actions) resulting in potential losses. It is totally dependent on the experience and ability of analysts to construct an information model called Control Structures, upon which analysts try to identify unsafe controls between system components. This paper proposes a formal approach to support the manual identification of UCAs, effectively and systematically. It allows analysts to mechanically extract Process Model, an important element that makes up the Control Structures, from a formal requirements specification for a software controller. It then concisely constructs the contents of Context Tables, from which analysts can identify all relevant UCAs effectively, using a software fault tree analysis technique. The case study with a preliminary version of a Korean nuclear reactor protections system shows the proposed approach's effectiveness and applicability.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3010-3016
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• 2023

In recent years, the development of radioactive waste containers for nuclear facility decommissioning and dismantling is a critical issue because the Taiwan domestic boiling water reactor nuclear power plant is going to be decommissioned. The main purpose of this research is to design a metal container that meets the structural requirements of related regulations. At first, the shielding analysis was performed by varying dimensions of radioactive waste to determine the storage efficiency of the container. Then, a series of structural analyses for operational and accidental conditions of the container with full load were conducted, such as lifting, stacking, and drop impact conditions. On the other hand, the field drop impact tests were carried out to ensure structural integrity. The present research demonstrates the structural safety of the developed container for decommissioned nuclear facilities in Taiwan.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4231-4235
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• 2022

Considering the importance of deuterium in nuclear science including medical and industrial researches such as (BNCT) and nuclear reactors respectively, it is important to study various possible ways in addition to common methods for measuring its concentration. This study is an effort to measure deuterium concentration using PGNAA. The main idea is to calculate the area under 2.23 MeV gamma-rays photo peak resulting from neutron collision with Hydrogen atoms which are in mix with deuterium in samples. The study carried out by both simulation and experiment. Monte Carlo MCNPX2.6 code has been used for simulation and based on its acceptable results an experimental setup has been arranged. The coordination of results was in the range of R = 0.99 and R = 0.98 in simulation and experiment respectively. The accuracy of the study has been investigated by measuring the concentration of an unknown sample by both PGNAA and Fourier transform infrared spectroscopy (FT-IR) methods in which there were acceptable correlation between these two methods.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1874-1879
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• 2024

Owing to strict radiation safety management in Korean nuclear power plants (NPPs), most radiation workers receive very low radiation doses, even lower than the annual dose limit for the general public. However, the occupational dose distribution indicates that some Korean NPP workers receive a relatively higher dose than the average dose. This inequity in radiation exposure could be reduced by providing customized radiation protection measures, such as dose constraints, to workers receiving relatively higher doses. In this study, dose normalization was performed to identify the highest radiation exposure work in Korean pressurized water reactors (PWRs). The results show that most of the occupational exposure in Korean PWRs occurs during the planned maintenance period. Finally, the three highest radiation exposure tasks in Korean PWRs were identified: nozzle dam installation and removal, eddy current testing, and man-way opening and closing.

• Journal of Animal Science and Technology
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• v.50 no.4
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• pp.561-572
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• 2008

Responses of real-time control parameters, such as ORP, DO and pH, to the conditions of biological animal wastewater treatment process were examined to evaluate the stability of real-time control using each parameter. Also an optimum index for supplemental carbon source addition based on NOx-N level was determined under a consideration of denitrification rate by endogenous respiration of microorganism and residual organic matter in liquor. Experiment was performed with lab-scale sequencing batch reactor(SBR) and working volume of the process was 45L. The distinctive nitrogen break point(NBP) on ORP-and DO-time profiles, which mean the termination of nitrification, started disappearing with the maintenance of low NH4-N loading rate. Also the NBP on ORP-and DO-time profiles was no longer observed when high NOx-N was loaded into the reactor, and the sensitivity of ORP became dull with the increase of NOx-N level. However, the distinctive NBP was constantly occurred on pH(mV)-time profile, maintaining unique profile patterns. This stable occurrence of NBP on pH(mV)-time profile was lasted even at very high NOx-N:NH4-N ratio(over 80:1) in reactor, and the specific point could be easily detected by tracking moving slope change(MSC) of the curve. Revelation of NBP on pH(mV)-time profile and recognition of the realtime control point using MSC were stable at a condition of over 300mg/L NOx-N level in reactor. The occurrence of distinctive NBP was persistent on pH(mV)-time profile even at a level of 10,000mg/L STOC(soluble total organic carbon) and the recognition of NBP was feasible by tracing MSC, but that point on ORP and DO-time profiles began to disappear with the increase of STOC level in reactor. The denitrfication rate by endogenous respiration and residual organic matter was about 0.4mg/L.hr., and it was found that 0.83 would be accepted as an index for supplemental carbon source addition when 0.1 of safety factor was applied.