• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3478-3487
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• 2022

In this study, we incorporate an anisotropic scattering scheme involving spherical harmonics into the method of characteristics (MOC). The neutron transport solution in a light water reactor can be significantly improved because of the impact of an anisotropic scattering source with the MOC flat source approximation. Several problems are selected to verify the proposed scheme and investigate its effects and accuracy. The MOC anisotropic scattering source is based on the expansion of spherical harmonics with Legendre polynomial functions. The angular flux, scattering source, and cross section are expanded in terms of the surface spherical harmonics. Later, the polynomial is expanded to achieve the odd and even parity of the source components. Ultimately, the MOC angular and scalar fluxes are calculated from a combination of two sources. This paper presents various numerical examples that represent the hot and cold conditions of a reactor core with boron concentration, burnable absorbers, and control rod materials, with and without a reflector or baffle. Moreover, a small critical core problem is considered which involves significant neutron leakage at room temperature. We demonstrate that an anisotropic scattering source significantly improves solution accuracy for the small core high-leakage problem, as well as for practical large core analyses.

• The Journal of Engineering Geology
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• v.25 no.2
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• pp.281-290
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• 2015

Log density data were collected and compared with the core density data throughout the southern Korean Peninsula. The comparison reveals that the log densities obtained from gamma-gamma log are much lower than the core densities obtained from laboratory density measurement of core samples. The anomalously low log densities can be attributed to the small-source density log data. Correlation analysis reveals differences between densities derived from the two methods, indicating that a data quality problem arises when using small-source log data. The problem is probably due to the fact that small-source data have not been obtained under ideal conditions for maintaining the appropriate relationship between gamma response and formation density. The frequency distribution characteristics of formation density in the southern Korean Peninsula could be determined using the core and the standard-source log data which are well-correlated.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.467-472
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• 1996

This study discusses special aspects of severe accident related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor.$^{1, 2)}$ The analytical comparison of three elements core to former two elements case is conducted including evaluation of suitable nuclear cross-section sets to account for the effects of system configulation, fuel and moderator mixture temperature, material dispersion and the other thermal-hydraulics. Three elements core ANS reactor is the alternative core design which was proposed as a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium (former uranium fuel is the baseline design value of 93%) A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies still on geometry, material constituents, and thermal-hydraulic conditions are verified. Therefore, the concepts of mitigative design features are qualified.d.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2460-2470
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• 2020

To study the thermophysical and neutronic properties of thorium-plutonium fuel, a conceptual design of a hybrid facility consisting of a subcritical Th-Pu reactor core and a source of additional D-D neutrons that places on the axis of the core is proposed. The source of such neutrons is a column of high-temperature plasma held in a long magnetic trap for D-D fusionreactions. This article presents computer simulation results of generation of thermonuclear neutrons in the plasma, facility neutronic properties and the evolution of a fuel nuclide composition in the reactor core. Simulations were performed for an axis-symmetric radially profiled reactor core consisting of zones with various nuclear fuel composition. Such reactor core containing a continuously operating stationary D-D neutron source with a yield intensity of Y = 2 × 10¹⁶ neutrons per second can operate as a nuclear hybrid system at its effective coefficient of neutron multiplication 0.95-0.99. Options are proposed for optimizing plasma parameters to increase the neutron yield in order to compensate the effective multiplication factor decreasing and plant power in a long operating cycle (3000-day duration). The obtained simulation results demonstrate the possibility of organizing the stable operation of the proposed hybrid 'fusion-fission' facility.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.07a
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• pp.160-163
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• 2002

Silica waveguide for an integrated diffractive optical head system was designed and fabricated. The waveguide was designed to optimize the optical efficiency of red and/or blue laser source, and a lab-made RF magnetron sputter was adopted to deposit silica cladding and core layers on SiO$_2$/Si substrates. The cladding and core layers were formed using commercial targets, and the former was done with ＃7740 and the latter with BK7 and BAK4, respectively The surface roughness of the waveguide layers was measured to be 30.3${\AA}$ for BK7 and 17.8${\AA}$ for BAK4, and the difference of refractive indices between core and cladding layers was 0.9% and 2.5%, respectively. The waveguide fabricated with the core layer of BK7 showed better optical properties when the final diffractive optical probe heads were measured with red laser(650nm) source.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1291-1302
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• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• Proceedings of the Korean Information Science Society Conference
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• 2002.04a
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• pp.214-216
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• 2002

본 논문에서는 many-to-many IP 멀티캐스팅을 위한 효율적인 Backbone Core Tree(BCT)생성 알고리즘에 대하여 제안한다. 본 논문의 제안기법은 Core Based Tree(CBT)에 기반을 두고 있다. CBT는 공유 트리를 이용하여 멀티캐스트 자료를 전달하기 때문에 Source Based Tree에 비하여 각 라우터가 유지해야 하는 상태 정보의 양에 적고 적용하기 간단하지만, Core 라우터 선택의 어려움과 트래픽이 Core로 집중되는 문제점을 가지고 있다. 이에 대한 보완책으로 Backbone Core Tree기법이 제안되었는데, 본 논문에서는 주어진 네트워크 위상 그래프에서 최소신장 트리를 만들고, 센트로이드를 이용하여 효율적인 BCT를 생성하는 알고리즘을 제안한다.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• The Journal of Asian Finance, Economics and Business
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• v.8 no.3
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• pp.967-976
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• 2021

This study aimed to investigate the mechanism by which service companies transfer their services overseas and create new value while interacting with local characteristics. A narrative analysis method was used in a case study of Ajisen Ramen, a Japanese service company that created a Japanese-style ramen restaurant, which experienced rapid growth in China. This study analyzed the restaurant as global service innovation and constructed a causal mechanism to explain the resulting rapid growth. In the pre-entry stage, the tangible value source core service facilitated its successful overseas transfer. In the post-entry stage, value source core service standardization and value sharing were interrelated and locally accepted factors. Knowledge of the local characteristics influenced the local storefront format that allowed contact with local customers. This local front format made it possible to offer local customers new value. The global service innovation mechanism developed through this study reflects a causal diagram that correlated the theoretical concepts of these events.