• Journal of the Korean Society for Aviation and Aeronautics
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• v.21 no.4
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• pp.47-52
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• 2013

In this paper, limit cycle flutter induced by Hopf bifurcation is studied with nonlinear system analysis approach and observed for the critical slowing down phenomenon. Considering an attractor of the dynamics of a system, when a small perturbation is applied to the system, the dynamics converge toward the attractor at some rate. The critical slowing down means that this recovery rate approaches zero as a parameter of the system varies and the size of the basin of attraction shrinks to nil. Consequently, in the pre-bifurcation regime, the recovery rates decrease as the system approaches the bifurcation. This phenomenon is one of the features used to forecast bifurcation before they actually occur. Therefore, studying the critical slowing down for limit cycle flutter behavior would have potential applicability for forecasting those types of flutter. Herein, modeling and nonlinear system analysis of the 2D airfoil with torsional nonlinearity have been discussed, followed by observation of the critical slowing down phenomenon.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1084-1096
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• 2023

On-the-fly(OTF) generation of Spectral Superhomogenization(SSPH) factors is analyzed in the multigroup(MG) whole core calculation employing pin-wise continuous energy(CE) slowing-down solutions. The motivation for the work is to avoid the huge computing time required for the generation of a parametrized SSPH factor library(PSSL) which is used to resolve the angular dependency of MG resonance cross sections, and also to exploit the advantage of flexible choice of a MG structure by using CE slowing-down solutions. Two pin-wise CE slowing-down methods, the equivalent Dancoff cell method and the shadowing effect correction method, are evaluated with the OTF SSPH method. The effectiveness of the OTF SSPH method is examined for various simplified and realistic core problems with various MG structures. It is demonstrated that the computing time overhead of this method is negligible whereas the solution accuracy is considerably enhanced.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.837-846
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• 2014

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.

• Journal of Radiation Protection and Research
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• v.10 no.2
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• pp.89-95
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• 1985

An alternative estimator for dose equivalent was derived. The original LET distribution concept was transformed into a charged particle fluence spectrum concept along with the definition of an average quality factor named slowing-down averaged quality factor by adopting the continuous slowing down approximation. With the alternative estimator, the dose equivalent delivered into a receptor located in a given radiation field can be directly and conveniently estimated in a Monte Carlo procedure. The slowing-down averaged quality factors for the energy range below 10 MeV were evaluated and tabulated for the charged particles which may be generated from the interactions of neutron with the nuclei composing soft tissue.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.182-189
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• 2018

A neutron generation system was developed to induce fissile fission in a lead slowing down spectrometer (LSDS) system. The source neutron is one of the key factors for LSDS system work. The LSDS was developed to quantify the isotopic contents of fissile materials in spent nuclear fuel and recycled fuel. The source neutron is produced at a multilayered target by the (e,${\gamma}$)(${\gamma}$,n) reaction and slowed down at the lead medium. Activation analysis of the target materials is necessary to estimate the lifetime, durability, and safety of the target system. The CINDER90 code was used for the activation analysis, and it can involve three-dimensional geometry, position dependent neutron flux, and multigroup cross-section libraries. Several sensitivity calculations for a metal target with different geometries, materials, and coolants were done to achieve a high neutron generation rate and a low activation characteristic. Based on the results of the activation analysis, tantalum was chosen as a target material due to its better activation characteristics, and helium gas was suggested as a coolant. In addition, activation in a lead medium was performed. After a distance of 55 cm from the lead surface to the neutron incidence, the neutron intensity dramatically decreased; this result indicates very low activation.

• Journal of the Korean Society of Radiology
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• v.4 no.4
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• pp.29-32
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• 2010

The neutron capture cross section of $^{197}Au$ has been measured relative to the $^{10}B(n,{\gamma})$ standard cross section by the neutron time-of-flight(TOF) method using a 46-MeV electron linear accelerator(linac) at the Research Reactor Institute, Kyoto University(KURRI). In order to experimentally prove the result obtained, the supplementary cross section measurement has been made from 0.1 eV to 10 keV using the Kyoto University Lead slowing-down spectrometer (KULS) coupling to the linac. The relative measurement by the TOF method has been normalized to the reference value(24.5 b) at 1 eV. The evaluated capture cross sections in JENDL/D-99 Dosimetry have been compared with the current measurements by the KULS experiments.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1199-1210
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• 2017

A new in-line method for sampling neutrons emitted in (${\alpha}$,n) reactions based on alpha particle source information has been developed for continuous-energy Monte Carlo simulations. The new method uses a continuous-slowing-down model coupled with (${\alpha}$,n) cross section data to precompute the expected neutron yield over the alpha particle lifetime. This eliminates the complexity and computational cost associated with explicit charged particle transport. When combined with an integrated alpha particle decay source sampling capability, the proposed method provides an efficient and accurate method for sampling (${\alpha}$,n) neutrons based solely on nuclide inventories in the problem, with no additional user input required. Results from several example calculations show that the proposed method reproduces the (${\alpha}$,n) neutron yields and energy spectra from reference experiments and calculations.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.380-387
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• 2015

A lead slowing-down spectrometer (LSDS) system is a promising nondestructive assay technique that enables a quantitative measurement of the isotopic contents of major fissile isotopes in spent nuclear fuel and its pyroprocessing counterparts, such as $^{235}U$, $^{239}Pu$, $^{241}Pu$, and, potentially, minor actinides. The LSDS system currently under development at the Korea Atomic Energy Research Institute (Daejeon, Korea) is planned to utilize a high-flux ($>10^{12}n/cm^2{\cdot}s$) neutron source comprised of a high-energy (30 MeV)/high-current (~2 A) electron beam and a heavy metal target, which results in a very intense and complex radiation field for the facility, thus demanding structural shielding to guarantee the safety. Optimization of the structural shielding design was conducted using MCNPX for neutron dose rate evaluation of several representative hypothetical designs. In order to satisfy the construction cost and neutron attenuation capability of the facility, while simultaneously achieving the aimed dose rate limit (< $0.06{\mu}Sv/h$), a few shielding materials [high-density polyethylene (HDPE)eBorax, $B_4C$, and $Li_2CO_3$] were considered for the main neutron absorber layer, which is encapsulated within the double-sided concrete wall. The MCNP simulation indicated that HDPE-Borax is the most efficient among the aforementioned candidate materials, and the combined thickness of the shielding layers should exceed 100 cm to satisfy the dose limit on the outside surface of the shielding wall of the facility when limiting the thickness of the HDPE-Borax intermediate layer to below 5 cm. However, the shielding wall must include the instrumentation and installation holes for the LSDS system. The radiation leakage through the holes was substantially mitigated by adopting a zigzag-shape with concrete covers on both sides. The suggested optimized design of the shielding structure satisfies the dose rate limit and can be used for the construction of a facility in the near future.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.17-25
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• 1993

Beta ray spectra of $^3H,\;^{14}C\;and\;^{36}Cl$ in liquid scintillation counting system have been calculated using the Monte Carlo method by which physical behaviors of particle transport in medium were simulated. The calculations have been carried out on the basis of beta rays being slowing down according to the continuous slowing down approximation(CSDA) model. Beta rays generated in simulation geometry were traced until they lost their energy below 0.3keV that in known to be the detection limit in the liquid scintillation counter. Scintillator solution in which pure beta emitting radionuclides were dissolved uniformly was assumed to be bottled in the shape of right circular cylinder with 12.5mm in radius and 35mm in height. The comparison of the calculated and measured results showed satisfactory agreement between those two, with slight discrepancy due to self quenching in the case of lower energy of emitted beta particles in the solution.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.549-555
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• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.