• Nuclear Engineering and Technology
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• 제51권5호
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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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• 제50권8호
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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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• 제52권12호
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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.

• 로봇학회논문지
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• 제17권2호
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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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• 제54권5호
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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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• 제55권8호
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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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• 제54권11호
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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.

• Journal of Animal Science and Technology
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• 제50권4호
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• pp.561-572
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• 2008

다양한 조건하에서 가축분뇨처리공정을 운전하면서 각 자동제어 인자의 반응을 분석하고 ORP, DO, pH(mV)-time profile를 이용한 자동제어 신뢰성을 평가하였다. 또한 무산소 조건에서의 잔존 유기물 및 미생물 자기산화에 의한 탈질율을 고려한 적정 외부탄소원 공급량 지표를 파악하였다. 실험은 45L의 유효용적을 지닌 실험실 규모의 SBR 공정을 이용하여 수행되었다. ORP-와 pH(mV)-, DO-time profile 상에서 완전질산화를 의미하는 NBP가 뚜렷하게 발현하던 중 NH4-N의 낮은 부하와 고농도 NOx-N 함유 폐수의 유입 및 불충분한 무산소 조건 제공이 이루어졌을 때 ORP-와 DO-time profile 상에서 NBP가 사라지기 시작하였으며 NOx-N의 지속적인 증가에 의해 ORP 값의 민감성이 둔화되기 시작하였다. 그러나 pH(mV)-time profile은 항상 일정한 변화패턴을 유지하면서 암모니아성 질소의 완전 질산화가 이루어졌을 때 뚜렷한 NBP를 발현하였다. NOx-N/NH4-N의 비가 80:1 수준까지 높아지는 조건하에서도 pH(mV)- time profile상에서의 이러한 안정적 NBP의 발현은 지속되었으며 발현되는 NBP는 MSC(Moving Slope Change)의 변화 패턴을 추적함에 의해 인식되도록 프로그램 할 수 있었다. pH(mV)-time profile에서의 NBP의 발현과 MSC를 이용한 자동제어시점 인식은 반응조내 NOx-N 농도가 무려 300mg/L 이상의 수준에서도 안정적이었다. 유기물 농도에 따른 자동제어 인자의 반응을 분석한 시험에서도 반응조내 유기물의 농도가 STOC 기준 약 10,000mg/L 수준으로 증가함에도 불구하고 pH(mV)-time profile 상에서의 이러한 NBP 발현은 지속되었으며 고농도 유기물 축적 하에서도 동일한 자동제어 알고리즘이 이용될 수 있음을 알 수 있었다. 잔존 유기물과 미생물 자기산화에 의한 탈질율은 약 0.4mg/L.hr로 분석되었으며 안전지수 0.1을 도입하여 산출된 NOx-N 기준 적정 외부탄소원 공급량은 0.83 STOC/NOx-N으로 파악되었다.

• Corrosion Science and Technology
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• 제12권2호
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• pp.102-112
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• 2013

A very high-temperature gas reactor (VHTR) is one of the next generation nuclear reactors owing to its safety, high energy efficiency, and proliferation-resistance. Heat is transferred from the primary helium loop to the secondary helium loop through an intermediate heat exchanger (IHX). Under VHTR environment Alloy 617 is being considered a candidate Ni-based superalloy for the IHX of a VHTR, owing to its good creep resistance, phase stability and corrosion resistance at high temperature. In this study, high-temperature corrosion tests were carried out at 850 - $950^{\circ}C$ in air and impure helium environments. Alloy 617 specimens showed a parabolic oxidation behavior for all temperatures and environments. The activation energy for oxidation was 154 kJ/mol in helium environment, and 261 kJ/mol in an air environment. The scanning electron microscope (SEM) and energy-dispersive x-ray spectroscopy (EDS) results revealed that there were a Cr-rich surface oxide layer, Al-rich internal oxides and depletion of grain boundary carbide after corrosion test. The thickness and depths of degraded layers also showed a parabolic relationship with the time. A corrosion rate of $950^{\circ}C$ in impure helium was higher than that in an air environment, caused by difference in the outer oxide morphology.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.