• Journal of the Korean Magnetics Society
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• v.15 no.2
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• pp.142-147
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• 2005

The embrittlement of fast neutron-irradiated reactor pressure vessel (RPV) steels was investigated by X-ray diffraction patterns at room temperature and $M\ddot{o}ssbauer$ spectroscopy at room- and liquid nitrogen-temperature. Neutron fluence on the samples were $10^{12},\;10^{13},\;10^{14},\;10^{15},\;10^{16},\;10^{17},\;10^{18}\;n/cm^2$. The X-ray diffraction patterns showed that the structure of the neutron unirradiated sample was bcc type, where as but the neutron irradiated samples with the fluence higher than $10^{17}\;n/{\cal}cm^2$ were so severely damaged, that bcc type structure disappeared. The $M\ddot{o}ssbauer$ spectra of all samples showed superposition of two or more sextets. In this paper all $M\ddot{o}ssbauer$ spectra were fitted by three set of sextet. The isomer shift and quadrupole splitting values were found around zero. At liquid nitrogen temperature, magnetic hyperfine field and absorption area increase rapidly S 1 sextet in the samples of $10^{17}\~10^{18}\;n/{\cal}cm^2$ neutron fluences. And at room temperature, magnetic hyperfine field and absorption increased rapidly at SI sextet in the samples of $10^{17}\~10^{18}\;n/{\cal}cm^2$ neutron fluences. This rapid increase of magnetic hyperfine field and absorption area were inferred to be caused by the change of $^{56}Fe,\;^{55}Mn$ into $^{57}Fe$ due to by neutron irradiation.

• The Journal of the Korea Contents Association
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• v.4 no.2
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• pp.76-82
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• 2004

The thermal neutron capture cross-section (at 2,200 m/s value) of the $^{141}$Pr(n,$\gamma$)$^{142}$Pr reaction was measured by an activation method by using the heavy water ($D_2$O) thermal neutron facility at the KUR(Kyoto University Reactor). The thermal neutron fiux used in this experiment was monitored with the$^{197}$Au(n,$\gamma$)$^{198}$Au standard cross-section. The previous results and the evaluated data of JENDL-3.2, ENDF/B-VI, and JEF-2.2 were in good agreement with the current result.

• Nuclear Engineering and Technology
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• v.19 no.4
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• pp.317-324
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• 1987

The purpose of this work is to present a spatial neutron kinetics computational scheme for the analysis of space-dependent transients like rod ejection accident of pressurized water reactors. In this work modified Borresen's 1.5 group coarse mesh scheme was formulated for the neutronic computation of the space-dependent transients and applied to the analysis of hypothetical rod ejection accident of KNU no. 1 PWR core at BOC, HZP. The computational accuracy of the modified Borresen's scheme is examined by comparing calculations for core power and control rod worths with startup core physics test results. Effects of such parameters as ejected rod worths and number of delayed neution group ell transient results as well as computational efficiency are also examined. OB this basis it is suggested that the modified Borresen's method is a useful scheme for the analysis of spatial neutron transients of PWR's.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1626-1637
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• 2020

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors. This ambitious benchmark aims to determine the uncertainty in light water reactors systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. The Gesellschaft für Anlagen und Reaktorsicherheit methodology is used to propagate the thermal-hydraulic uncertainty of macroscopic parameters through TRACE5.0p3/PARCSv3.0 coupled code. The main innovative points achieved in this work are i) a new thermal-hydraulic model is developed with a highly-accurate 3D core discretization plus an iterative process is presented to adjust the 3D bypass flow, ii) a control rod insertion occurrence -which data is obtained from a real PWR test- is used as a transient simulation, iii) two approaches are used for the propagation process: maximum response where the uncertainty and sensitivity analysis is performed for the maximum absolute response and index dependent where the uncertainty and sensitivity analysis is performed at each time step, and iv) RESTING MATLAB code is developed to automate the model generation process and, then, propagate the thermal-hydraulic uncertainty. The input uncertainty information is found in related literature or, if not found, defined based on expert judgment. This paper, first, presents the Gesellschaft für Anlagen und Reaktorsicherheit methodology to propagate the uncertainty in thermal-hydraulic macroscopic parameters and, then, shows the results when the methodology is applied to a PWR reactor.

• Progress in Superconductivity and Cryogenics
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• v.16 no.1
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• pp.9-13
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• 2014

Effects of neutron irradiation on the superconducting properties of the undoped $MgB_2$ and the carbon(C)-doped $MgB_2$ bulk superconductors, prepared by an in situ reaction process using Mg and B powder, were investigated. The prepared $MgB_2$ samples were neutron-irradiated at the neutron fluence of $10^{16}-10^{18}n/cm^2$ in a Hanaro nuclear reactor of KAERI involving both fast and thermal neutron. The magnetic moment-temperature (M-T) and magnetization-magnetic field (M-H) curves before/after irradiation were obtained using magnetic property measurement system (MPMS). The superconducting critical temperature ($T_c$) and transition width were estimated from the M-T curves and critical current density ($J_c$) was estimated from the M-H curves using a Bean's critical model. The $T_cs$ of the undoped $MgB_2$ and C-doped $MgB_2$ before irradiation were 36.9-37.0 K and 36.6-36.8 K, respectively. The $T_cs$ decreased to 33.2 K and 31.6 K, respectively after irradiation at neutron fluence of $7.16{\times}10^{17}n/cm^2$, and decreased to 22.6 K and 24.0 K, respectively, at $3.13{\times}10^{18}n/cm^2$. The $J_c$ cross-over was observed at the high magnetic field of 5.2 T for the undoped $MgB_2$ irradiated at $7.16{\times}10^{17}n/cm^2$. The $T_c$ and $J_c$ variation after the neutron irradiation at various neutron fluences were explained in terms of the defect formation in the superconducting matrix by neutron irradiation.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1120-1130
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• 2020

The CREOLE experiment performed In the EOLE critical facility located In the Nuclear Center of CADARACHE - CEA have allowed us to get interesting and complete experimental information on the temperature effects in the light water reactor lattices. To analyze these experiments with accuracy an elaborate calculation scheme using the Monte Carlo method implemented in the MCNP6.1 code and the ENDF/B-VII.1 cross section library has been developed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original library. The main purpose of this analysis is the qualification of the ENDF/B-VII.1 nuclear data for the prediction of the Reactivity Temperature Coefficient while ensuring the ability of the MCNP6.1 system to model such a complex experiment as CREOLE. We have analyzed the case of UO₂ lattice with 1166 ppm of boron in ordinary water moderator in specified temperatures. A detailed comparison of the calculated effective multiplication factors with the reference ones [1] in room temperature presented in this work shows a good agreement demonstrating the validation of our 3D calculation model. The discrepancies between calculations and the differential measurements of the Reactivity Temperature Coefficient for the analyzed configuration are relatively small: the maximum discrepancy doesn't exceed 1,1 pcm/℃. In addition to the analysis of direct differential measurements of the reactivity temperature coefficient performed in the poisoned UO₂ lattice configuration, we have also analyzed integral measurements in UO₂ clean lattice configuration using equivalency of the integral temperature reactivity worth with the driver core fuel reactivity worth and soluble boron reactivity worth. In this case both of the ENDF/B-VII.1 and JENDL.4 libraries were used in our analysis and the obtained results are very similar.

• Progress in Medical Physics
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• v.9 no.1
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• pp.55-64
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• 1998

In recent, the demand of development of the high dose rate brachytherapy source increased for substitute for Co-60 source by iridium source, since the supplying Co-60 source is very depressed and the high dose rate brachytherapy sources are entirely imported from the abroad. This study investigated the exposure rates and isotropic dose distributions for the Ir-192 source produced from $\^$191/Ir(n,r)$\^$192/Ir by nuclear reactor in Korea Atomic Energy Research Institute. The activity of source was obtained an 1.012 Ci (the initial activity without encapsulation was 2,87Ci) by measurement with encapsuled stainless steel. The exposure rate of provided Ir-192 source was determined on 6.36 ${\pm}$ 0.147 Rm$^2$/h-GBq (2.350 ${\pm}$ 0.054 Rcm$^2$/mCi-hr) within ${\pm}$ 2.2% discrepancy with IC-10 ion chamber (0.14 cc) which was mounted on the acrylic jig to 5, 10 and 20 cm from the center of source. The calculated doses with 22 most significant spectrum lines were corrected with intrinsic efficiency of the germanium detector were compared to measured exposure dose rates within ${\pm}$3.8 % discrepancy. The authors confirmed the high dose rate Ir-192 source could be replaced the long decayed Co-60 source via investigation of the isotropic dose distributions in lateral, source axis and diagonal direction of source center are very closed to within 3% uncertainties. Especially, this exposure rate constant and isotropic dose distribution will be fundamental to build the high dose rate source and develop the computed therapy planning system.

• Progress in Medical Physics
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• v.24 no.3
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• pp.171-175
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• 2013

Applicator of various kind of number ten kinds is used to raise from efficiency of brachytherapy to maximum. The compatibility of radiation source and applicator is very important subject for safety brachytherapy. Developed high dose rate brachytherapy source through Hanaro nuclear reactor in Korea Atomic Energy Research Institute and improve compatibility with using equipment in present. In this research, we wished to evaluate stability mechanical safety of radiation source and we developed phantom for evaluate several quality about Ir-192 sealed source that improve newly in Korea Atomic Energy Research Institute and is improved. The result for suitability of Ir-192 HDR source with brachytherapy tools that did normal operation in 2.2~2.7 cm extent about change of equal curvature and consider change of sudden curvature that did normal operation in radius 1.5~1.8 cm extent.

• Nuclear Engineering and Technology
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• v.1 no.2
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• pp.91-101
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• 1969

Resistivity changes of n-type float-zone silicon crystals with 6.4$\times$10$^{14}$ to 1.25$\times$10$^{17}$ phosphorus atoms/㎤ due to irradiation by (1) 1 MeV electrons, (2) two types of research reactors, and (3) $Co^{60}$ ${\gamma}$-ray sources were investigated. The results were analyzed on the basis of a simple exponential formula derived by Buehler. While the formula gave a fair fit in the low fluence range in most cases, the deviation was quite appreciable in the case of 1 MeV electron irradiation, and a linear change gave better fit in some cases. The large change in the carrier removal rate in electron-irradiated samples in the high fluence range was analyzed in detail in terms of the Fermi level cross-over of the defect levels. Based on the damage constants evaluated from the initial portion of data where the formula was applicable, the relative effectiveness of various radiation sources in causing the resistivity change in n-type silicon was compared. The TRIGA Mark II reactor neutrons, for example, were found to be about 40 times more effective than 1 MeV electrons. The dependence of the damage constant on the initial carrier concentration was also examined. The physical basis of the exponential law and the effect of the Fermi level cross-over of the defect levels on the resistivity change in the high fluence ranges are discussed.