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RISS 인기검색어
- 검색키워드
- 저자 : Byoung-Uhn (검색결과 33 건)
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Byoung-Uhn Bae,Seok Cho,Jae Bong Lee,Yu-Sun Park,Jongrok Kim,Kyoung-Ho Kang Korean Nuclear Society 2023 Nuclear Engineering and Technology Vol.55 No.7
To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.
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Bae, Byoung-Uhn,Lee, Jae Bong,Park, Yu-Sun,Kim, Jongrok,Kang, Kyoung-Ho Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.8
To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.
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Bae, Byoung-Uhn,Yun, Byong-Jo,Yoon, Han-Young,Song, Chul-Hwa,Park, Goon-Cherl Elsevier 2010 Nuclear engineering and design Vol.240 No.9
<P><B>Abstract</B></P><P>To enhance the multi-dimensional analysis capability for a subcooled boiling two-phase flow, the one-group interfacial area transport equation was improved with a source term for the bubble lift-off. It included the bubble lift-off diameter model and the lift-off frequency reduction factor model. The bubble lift-off diameter model took into account the bubble's sliding on a heated wall after its departure from a nucleate site, and the lift-off frequency reduction factor was derived by considering the coalescences of the sliding bubbles. To implement the model, EAGLE (elaborated analysis of gas–liquid evolution) code was developed for a multi-dimensional analysis of two-phase flow. The developed model and EAGLE code were validated with the experimental data of SUBO (subcooled boiling) and SNU (Seoul National University) test, where the subcooled boiling phenomena in a vertical annulus channel were observed. Locally measured two-phase flow parameters included a void fraction, interfacial area concentration, and bubble velocity. The results of the computational analysis revealed that the interfacial area transport equation with the bubble lift-off model showed a good agreement with the experimental results of SUBO and SNU. It demonstrates that the source term for the wall nucleation by considering a bubble sliding and lift-off mechanism enhanced the prediction capability for the multi-dimensional behavior of void fraction or interfacial area concentration in the subcooled boiling flow. From the point of view of the bubble velocity, the modeling of an increased turbulence induced by boiling bubbles at the heated wall enhanced the prediction capability of the code.</P>
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Bae, Byoung Uhn,Lee, Keo Hyoung,Kim, Yong Soo,Yun, Byong Jo,Park, Goon Cherl Elsevier 2008 Nuclear engineering and design Vol.238 No.9
<P><B>Abstract</B></P><P>A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal–hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal–hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal–hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.</P>
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Bae, Byoung-Uhn,Kim, Seok,Park, Yu-Sun,Kang, Kyoung-Ho,Ahn, Tae-Hwan,Yun, Byong-Jo Elsevier 2017 Annals of nuclear energy Vol.107 No.-
<P><B>Abstract</B></P> <P>PAFS (Passive Auxiliary Feedwater System) is one of the advanced safety features in the design of the APR+ (Advanced Power Reactor) nuclear power plant, in order to cool down the reactor coolant system without any external supply of electricity or water during an accident. The driving force of the PAFS is condensation inside horizontal tubes of PCHX (Passive Condensation Heat Exchanger). To improve the prediction capability for the condensation heat transfer in the PCHX, advanced wall condensation models which mechanistically consider the wall condensation with the different flow regimes in a horizontal channel were implemented into a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant). The models were evaluated by comparing to the experimental data from the PASCAL (PAFS Condensing Heat Removal Assessment Loop) test facility, which is a separate effect test to validate the cooling performance of the PCHX. Calculation results showed that the condensation model package (Ahn et al., 2014) presented the enhanced the prediction capability of the SPACE code in predicting the steam pressure and temperature under quasi-steady state conditions. Since this model package is composed of mechanistic models by considering the film condensation in the steam phase and the convection in the condensate liquid separately, it contributed to enhance the prediction capability of the SPACE code and reduce the conservatism in the prediction of the cooling capability of the PAFS. The SPACE code with improvement of the wall condensation model can be applied in predicting the cooling and operation capability of the PAFS more realistically in the safety analysis of APR+ nuclear power plant.</P> <P><B>Highlights</B></P> <P> <UL> <LI> This study focused on enhancing the prediction for the condensation phenomenon inside the PCHX. </LI> <LI> Advanced models for condensation were implemented into the SPACE code. </LI> <LI> Mechanistic modeling contributed to improve wall condensation model of the SPACE. </LI> </UL> </P>
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Bae, Byoung-Uhn,Kim, Seok,Park, Yu-Sun,Kang, Kyoung-Ho Elsevier 2020 Annals of nuclear energy Vol.139 No.-
<P><B>Abstract</B></P> <P>To provide a passive cooling system for the reactor containment, Passive Containment Cooling System (PCCS) was adopted in the design of i-POWER nuclear power plant. This study focused on validation tests for condensation heat transfer of the PCCS heat exchanger using the CLASSIC facility. The tests include investigation of the condensation heat transfer in prototypic single tube and bundle tubes. From the single tube experiments, condensation heat transfer model was proposed to reflect the PCCS heat exchanger tube geometry. Experimental results in the bundle tube show consistent trend compared to the proposed heat transfer model from the single tube test. The local condensation heat transfer coefficient of inside tubes was smaller than the average value due to a shadow effect by a larger mass fraction of non-condensable gas, so that design of the PCCS should take into account degradation of the condensation heat removal in the bundle geometry.</P> <P><B>Highlights</B></P> <P> <UL> <LI> This study focused on condensation heat transfer of the bundle heat exchanger for PCCS design. </LI> <LI> The local HTC of tubes located inside was smaller than the average due to a shadow effect. </LI> <LI> The PCCS design should consider the degradation of the condensation heat removal of inside tubes. </LI> </UL> </P>
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Bae, Byoung-Uhn,Park, Yu-Sun,Kim, Jong-Rok,Kang, Kyoung-Ho,Choi, Ki-Yong Elsevier 2018 Nuclear engineering and design Vol.328 No.-
<P><B>Abstract</B></P> <P>Station blackout (SBO) is one of the most important design extension conditions (DECs) because a total loss of the heat sink leads to a core uncovery or damage without any proper operator action. During a long transient of an SBO, a steam generator tube rupture (SGTR) accident can occur when a steam generator tube is exposed to a superheated steam flow. In this study, a prolonged SBO with an SGTR occurrence was experimentally investigated by performing an integral effect test with the ATLAS facility. For the transient simulation, the tube rupture was simulated when the core water level became below the top of the active core. As an accident management measure, the auxiliary feedwater was supplied to the intact steam generator when the maximum heater surface temperature started to show an excursion behavior. The experimental results showed that a single tube rupture could not sufficiently reduce the primary system pressure before the injection of the auxiliary feedwater. A delayed supply of the auxiliary feedwater after an excursion of the heater surface temperature successfully cooled the primary system until the end of the transient, where inflow of the coolant from the pressurizer contributed to a recovery of the coolant inventory in the core. The natural circulation flow characteristics in the primary system showed an oscillating behavior depending on the heat removal rate of the steam generators. This integral effect test data can be used to evaluate the prediction capability of safety analysis codes and identify code deficiencies in predicting an SBO transient with an SGTR occurrence.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A station blackout with a steam generator tube rupture was investigated in ATLAS. </LI> <LI> The multiple failure accident can be mitigated by the auxiliary feedwater supply. </LI> <LI> Sensitivity analysis showed a possibility of core damage in multiple tube rupture. </LI> </UL> </P>
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Experimental Study of Condensation Heat Exchanger in PAFS (Passive Auxiliary Feedwater System)
Seok Kim(김석),Byoung-Uhn Bae(배병언),Bok-Deuk Kim(김복득),Kyung Ho Kang(강경호),Byong-Jo Yun(윤병조) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.10
PAFS (Passive Auxiliary Feedwater System) is a passive cooling system on the secondary system of APR+ (Advanced Power Reactor Plus). It can replace the conventional active cooling system for auxiliary feedwater injection to the steam generator by a passive way, and it cools down the secondary system of the steam generator by heat transfer at the condensation heat exchanger installed in the PCCT (Passive Condensation Cooling Tank). To validate a cooling performance of PAFS, a separate effect test loop has been constructed at KAERI (Korea Atomic Energy Research Institute), which is named PASCAL (PAFS Condensing heat removal Assessment Loop). It simulates a single tube of the horizontal heat exchanger, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology. In this study, two-phase flow phenomena in a horizontal heat exchanger and PCCT (Passive Condensate Cooling Tank) for the facility were investigated and the cooling capability of the condensation heat exchanger was validated in the steady-state experiment. The experimental results in this study will contribute to validate a thermal hydraulic system analysis code or a CFD code for the multi-phase flow.
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