• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1181-1194
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• 2019

A new moving-mesh Finite Volume Method (FVM) for the efficient solution of the two-dimensional neutron diffusion equation is introduced. Many other moving-mesh methods developed to solve the neutron diffusion problems use a relatively large number of sophisticated mathematical equations, and so suffer from a significant complexity of mathematical calculations. In this study, the proposed method is formulated based on simple mathematical algebraic equations that enable an efficient mesh movement and CV deformation for using in practical nuclear reactor applications. Accordingly, a computational framework relying on a new moving-mesh FVM is introduced to efficiently distribute the meshes and deform the CVs in regions with high gradient variations of reactor power. These regions of interest are very important in the neutronic assessment of the nuclear reactors and accordingly, a higher accuracy of the power densities is required to be obtained. The accuracy, execution time and finally visual comparison of the proposed method comprehensively investigated and discussed for three different benchmark problems. The results all indicated a higher accuracy of the proposed method in comparison with the conventional fixed-mesh FVM.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2151-2161
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• 2021

Nuclear data (ND) sensitivity and uncertainty (S/U) quantification in shielding applications is performed using deterministic and probabilistic approaches. In this paper the validation of the newly developed deterministic program package ASUSD (ADVANTG + SUSD3D) is presented. ASUSD was developed with the aim of automating the process of ND S/U while retaining the computational efficiency of the deterministic approach to ND S/U analysis. The paper includes a detailed description of each of the programs contained within ASUSD, the computational workflow and validation results. ASUSD was validated on two shielding benchmark experiments from the Shielding Integral Benchmark Archive and Database (SINBAD) - the fission relevant ASPIS Iron 88 experiment and the fusion relevant Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) mock-up experiment. The validation process was performed in two stages. Firstly, the Denovo discrete ordinates transport solver was validated as a standalone solver. Secondly, the ASUSD program package as a whole was validated as a ND S/U analysis tool. Both stages of the validation process yielded excellent results, with a maximum difference of 17% in final uncertainties due to ND between ASUSD and the stochastic ND S/U approach. Based on these results, ASUSD has proven to be a user friendly and computationally efficient tool for deterministic ND S/U analysis of shielding geometries.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.165-169
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• 2018

An assessment for determining N-16 activity concentrations during the operation condition of Bangladesh Atomic Energy Commission TRIGA Research Reactor was performed employing several governing equations. The radionuclide N-16 is a high energy (6.13 MeV) gamma emitter which is predominately created by the fast neutron interaction with O-16 present in the reactor core water. During reactor operation at different power level, the concentration of N-16 at the reactor bay region may increase causing radiation risk to the reactor operating personnel or the general public. Concerning the safety of the research reactor, the present study deals with the estimation of N-16 activity concentrations in the regions of reactor core, reactor tank, and reactor bay at different reactor power levels under natural convection cooling mode. The estimated N-16 activity concentration values with 500 kW reactor power at the reactor core region was $7.40{\times}10^5Bq/cm^3$ and at the bay region was $3.39{\times}10^5Bq/cm^3$. At 3 MW reactor power with active forced convection cooling mode, the N-16 activity concentration in the decay tank exit water was also determined, and the value was $4.14{\times}10^{-1}Bq/cm^3$.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3398-3405
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• 2021

This study observed the effect of neutron irradiation and ageing on the microstructure, hardness, and corrosion resistance of SAV-1 (Al-Mg-Si) alloy. The investigated material was irradiated with neutrons to fluences of 10²¹-10²⁶ n/m² in the WWR-K research reactor and kept in dry storage. Long-term irradiation led to an increase in hardness of the alloy and a deterioration of pitting corrosion resistance. Post-irradiation ageing for 1 h at 100-300 ℃ resulted in a decrease in microhardness of the irradiated SAV-1. The effect of post-irradiation ageing on pitting corrosion was made clear through the formation of Guinier-Preston zones and secondary precipitates in the Al matrix. Ageing at 250 ℃ corresponded to the development of stable microstructure and the highest corrosion resistance for the irradiated samples. Mg₂Si, Si, and needle-shaped β" precipitates were formed in SAV-1 alloy that was irradiated with low fluences. β" and clusters of rod-shaped B-type precipitates were observed in highly irradiated samples. The precipitates were similar to those seen in non-irradiated pseudo-binary Al-Mg₂Si alloys with Si excess.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1243-1251
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• 2020

Irradiation-induced damage of binderless nanoporous-isotropic graphite (NPIG) prepared by isostatic pressing of mesophase carbon microspheres for molten salt reactor was investigated by 3.0 MeV He⁺ irradiation at room temperature and high temperature of 600 ℃, and IG-110 was used as the comparation. SEM, TEM, X-ray diffraction and Raman spectrum are used to characterize the irradiation effect and the influence of temperature on graphite radiation damage. After irradiation at room temperature, the surface morphology is rougher, the increase of defect clusters makes atom flour bend, the layer spacing increases, and the catalytic graphitization phenomenon of NPIG is observed. However, the density of defects in high temperature environment decreases and other changes are not obvious. Mechanical properties also change due to changes in defects. In addition, SEM and Raman spectra of the cross section show that cracks appear in the depth range of the maximum irradiation dose, and the defect density increases with the increase of irradiation dose.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1099-1109
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• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• Journal of Radiation Protection and Research
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• 제11권2호
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• pp.139-145
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• 1986

중성자와 우라늄의 핵반응에 의해 생성된 핵분열생성들의 물리적 특성을 이용하며 원자로 내의 핵연료 상태를 해석하는 모델을 개선하였다. 이 모델에서는 고체 핵연료 내에서 특정핵종의 핵분열 생성물의 생성과 이것이 원자로 냉각재까지 방출되는 과정을 계산하고 추적하여 방사능농도와 결함 핵연료봉의 수를 관계짓는 방정식의 계수들을 결정한다. 핵분열생성들의 거동은 이탈(knock out)과 이동(migration) 두 부분으로 나누어 해석하였으며 트램프 우라늄의 영향을 분리할 수 있도록 하였다. 실측자료로는 가압 경수형 원자로인 고리 원자력발전소 1호기의 1차 냉각재를 분석해서 얻은 I-131과 I-133의 방사능 강도를 이용하였다. 이 실험자료와 위 방정식에서 구한 방사능 강도로부터 구한 결함 핵연료의 수는 제 3 주기에서 $9.34{\pm}1.13$개 제 6 주기에서 $0.294{\pm}0.092$개로 나타났다.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1208-1212
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• 2020

Knowledge on fuel enthalpy and its temperature derivative, the heat capacity, are important quantities in determination of fuel behavior in normal reactor operation and reactor transients. The aim of this study is to compare the heat capacity of oxide and nitrite fuels by using Einstein-Debye approximation. A simple analytical expression was performed to calculate the heat capacity of fuels. To test the validity and reliability, the calculated formulas were compared to published results for various nuclear fuels including UO₂, ThO₂, PuO₂ and UN. Calculated formulas yielded results in consistent with literature.

• 한국정보디스플레이학회:학술대회논문집
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• 한국정보디스플레이학회 2002년도 International Meeting on Information Display
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• pp.163-166
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• 2002

We studied the growth of large-grain pentacene film by organic vapour phase deposition. The optimizations of the growth of pentacene are carried out by varying the gas pressure in the reactor and substrate temperature. We found that the grain size depends strongly on the gas pressure in the reactor. The grain size of $20{\mu}m$ has been obtained at the gas pressure of 200 Torr. The film was found to be strongly (001) oriented and its grain size decreases with decreasing the gas pressure.