• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2004년도 학술대회지
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• pp.131-137
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• 2004

This paper is intended as an investigation of the deposition characteristics and mass transport flux estimation in the Nakdong estuary. In order to understand the effects of the tidal current circulation which influenced to an estuary terrain change, the seawater circulation calculation by the use of 2D numerical model for the three cases of without riverflow, mean and flood riverflow quantity condition practiced and each sectional net-flux of water quantity between sandbars(so called, dung) estimated. It may be that an estuary terrain change due to the large scale construction and reclamation at the Nakdong estuary influence to the long-time deposition characteristics. by the revim for the old research, we know that the development of the local sandbars has been moved toward the east-side from the west-side estuary area after the construction of the Nakdong river dike, at present the strong-acted location is the Bakhap-dung of the front sea of Tadea. The seawater circulation pattern at this large scale area of tidal flat bring on a change due to the water quantity outflowing from the Nakdong river. Base on the calculated results for the section net-flux of water quantity, we see that the accumulating action very strong at the local sea around Jangjado, Bakhapdung and Tadae for the case of flood riverflow quantity condition, but at the local sea around Jinudo for the another cases. Consequently, it is emphasized that in the Nakdong estuary the main sensitive regions which influenced from the discharge of riverflow were the local sea around Jangjado, Bakhapdung, Tadae and Jinudo.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1853-1859
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• 2019

The two-thermocouple method was investigated experimentally to evaluate its accuracy for the measurement of local wall temperature and heat flux on a heat transfer tube with an electric heater rod installed in an annulus channel. This work revealed that a thermocouple flush-mounted in a surface groove serves as a good reference method for the accurate measurement of the wall temperature, whereas two thermocouples installed at different depths in the tube wall yield large bias errors in the calculation of local heat flux and wall temperature. These errors result from conductive and convective changes due to the fin effect of the thermocouple sheath. To eliminate the bias errors, we proposed a calibration method based on both the local heat flux and Reynolds number of the cooling water. The calibration method was validated with the measurement of local heat flux and wall temperature against experimental data obtained for single-phase convection and two-phase condensation flows inside the tube. In the manuscript, Section 1 introduces the importance of local heat flux and wall temperature measurement, Section 2 explains the experimental setup, and Section 3 provides the measured data, causes of measurement errors, and the developed calibration method.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.335-346
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• 2013

The Critical Heat Flux (CHF) of water with dispersed alumina nanoparticles was measured for the geometry and flow conditions relevant to the In-Vessel Retention (IVR) situation which can occur during core melting sequences in certain advanced Light Water Reactors (LWRs). CHF measurements were conducted in a flow boiling loop featuring a test section designed to be thermal-hydraulically similar to the vessel/insulation gap in the Westinghouse AP1000 plant. The effects of orientation angle, pressure, mass flux, fluid type, boiling time, surface material, and surface state were investigated. Results for water-based nanofluids with alumina nanoparticles (0.001% by volume) on stainless steel surface indicate an average 70% CHF enhancement with a range of 17% to 108% depending on the specific flow conditions expected for IVR. Experiments also indicate that only about thirty minutes of boiling time (which drives nanoparticle deposition) are needed to obtain substantial CHF enhancement with nanofluids.

• 한국해안·해양공학회논문집
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• 제28권3호
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• pp.160-170
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• 2016

본 연구에서는 조석 유입구에서의 단면통과 유량 및 조량 산정을 목적으로 목포해역 내 조석 유입구에서 ADCP를 이용한 단면 유속관측이 수행되었다. 우선, 원시 관측자료는 지배적인 유속성분 도출을 위하여 단면의 회전정도를 고려한 좌표회전이 수행되었으며, 정규화를 통하여 수평/수직방향 Sigma 좌표로 변환되었다. 또한, 관측이 이루어지지 않는 저면과 수면 근처의 Blank zone 데이터는 von-Karman 유속분포식을 이용하여 보간되었다. 보정된 자료를 이용하여 단면통과 유량이 정량적으로 산정되었는데, 조석에 따른 단면통과 유량의 변화양상 뿐만 아니라 목포해역의 낙조우세적 조석특성이 잘 나타나는 것으로 확인되었다. 또한, 과거 연구들에서 제시되는 조량 산정자료의 조석조건을 보완하여 현장 관측자료에 기반한 조량산정 방법의 조건이 재정립되었으며, 새로운 정의에 근거하여 조석 유입구에서의 단면 유속관측 자료를 이용한 조량의 정량적 산정이 국내 최초로 수행되었다. 조량 산정결과는 과거 연구들로부터 도출된 조량과 조석 유입구 흐름 단면적의 상관관계 결과와 비교 분석되었으며, 산정된 결과는 과거 연구결과들과 잘 일치하는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.281-290
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• 2014

In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

• Nuclear Engineering and Technology
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• 제42권3호
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• pp.297-304
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• 2010

In the design of advanced light water reactors (ALWRs) and in the safety assessment of currently operating nuclear power plants, it is necessary to evaluate the possibility of experiencing a degraded core accident and to develop innovative safety technologies in order to assure long-term debris cooling. The objective of this experimental study is to investigate the enhancement factors of dryout heat flux in debris beds by coolant injection from below. The experimental facility consists mainly of an induction heater, a double-wall quartz-tube test section containing a steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of the particle bed was achieved in the radial direction and the axial variation was within 20%. This paper reports the experimental data for 3.2 mm and 4.8 mm particle beds with a 300 mm bed height. The dryout heat density data were obtained for both the top-flooding and the forced coolant injection from below with an injection mass flux of up to $1.5\;kg/m^2s$. The dryout heat density increased as the rate of coolant injection increased. At a coolant injection mass flux of $1.0\;kg/m^2s$, the dryout heat density was ${\sim}6.5\;MW/m^3$ for the 4.8 mm particle bed and ${\sim}5.6\;MW/m^3$ for the 3.2 mm particle bed. The enhancement factors of the dryout heat density were 1.6-1.8.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3516-3525
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• 2022

The high capture cross-section (𝜎_c) of Gadolinium (Gd-155 and Gd-157) causes reactivity penalty and swing at the initial stage of fuel burnup in Pressurized Water Reactor (PWR). The present study is concerned with the feasibility of the combination of mixed burnable poison with both low and high 𝜎_c as an approach to minimize these effects. Two considered reference designs are fuel assemblies with 24 IBA rods of Gd₂O₃ and Er₂O₃ respectively. Models comprise nuclear fuel with a homogeneous mixture of Er₂O₃, AmO₂, SmO₂, and HfO₂ with Gd₂O₃ as well as the coating of PaO₂ and ZrB₂ on the Gd₂O₃ pellet's outer surface. The infinite multiplication factor was determined and reactivity was calculated considering 3% neutron leakage rate. All models except Er₂O₃ and SmO₂ showed expected results namely higher values of these parameters than the reference design of Gd₂O₃ at the early burnup period. The highest value was found for the model of PaO₂ and Gd₂O₃ followed by ZrB₂ and HfO₂. The cycle burnup, discharge burnup, and cycle length for three batch refueling were calculated using Linear Reactivity Model (LRM). The pin power distribution, energy-dependent neutron flux and Fuel Temperature Coefficient (FTC) were also studied. An optimization of model 1 was carried out to investigate effects of different isotopic compositions of Gd₂O₃ and absorber coating thickness.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.1-10
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• 2022

This study discusses the effect of using ²³²Th instead of ²³⁸U on the neutronic characteristics and the main operating parameters of the pressurized water reactor (PWR). MCNPX version 2.7 was used to compare the neutronic characteristics of UO₂ with (Th, ²³⁵U)O₂ and (Th, ²³³U) O₂. Firstly, the infinity multiplication factor (K_inf), thermal neutron flux, and power distribution have been studied for the investigated fuel types. Secondly, the effect of Gd₂O₃ and Er₂O₃ on the K_inf and on the radial thermal neutron flux and thermal power has been investigated to distinguish which of them is more suitable than the other in reactivity management. Thirdly, to illustrate the effectiveness of ²³²Th in decreasing the inventory of both the actinides and non-actinides, the concentration of plutonium (Pu) isotopes and minor actinides (MAs) has been simulated with the fuel burnup. Besides, due to their large thermal neutron absorption cross-section, the concentrations of ¹³⁵Xe, ¹⁴⁹Sm, and ¹⁵¹Sm with the fuel burnup have been investigated. Finally, the main safety parameters such as the reactivity worth of the control rods (ρ_CR), the effective delayed neutron fraction β_eff, and the Doppler reactivity coefficient (DRC) were calculated to determine to which extent these fuel types achieve the acceptable limits.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1287-1300
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• 2017

Based on the observation that ignoring the angle dependency of multigroup resonance cross sections within a fuel pellet would result in nontrivial underestimation of the spatial self-shielding of flux, a parametrized spectral superhomogenization (SPH) factor library (PSSL) method is developed as a practical means of resolving the problem. Region-wise spectral SPH factors are calculated by the normal and transport corrected SPH iterations after ultrafine group slowing down calculations over various light water reactor pin-cell configurations. The parametrization is done with fuel temperature, U-238 number density, fuel radius, moderator source represented by ${\Sigma}_{mod}V_{mod}$, and the number density ratio of resonance nuclides to that of U-238 in a form of resonance interference correction factors. The parametrization is successful in that the root mean square errors of the interpolated SPH factors over the fuel regions of various pin-cells are within 0.1%. The improvement in reactivity error of the PSSL method is shown to be superior to that by the original SPH method in that the reactivity bias of -200 pcm to -300 pcm vanishes almost completely. It is demonstrated that the environment effect takes only about 4% in the reactivity improvement so that the pin-cell based PSSL method is effective in the assembly problems.