• 토지주택연구
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• 제11권3호
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• pp.69-74
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• 2020

Since the design Standard for Energy Conservation in Building was implemented in 2008 for the first time, building elements such as window and door should satisfy the minimum criteria to apply for a building. Though its regulation does not cover the whole building yet, recent demand to reduce energy consumption in building sector grows rapidly year by year and also draws a lot of interest to ensure the whole building level. For example, a Zero energy building, one of low-energy buildings, requires a customized solution to resolve the air leakage issue to meet the standards in achieving the high level of air tightness. In this study, six non-residential buildings were tested by fan pressurization method to observe the air tightness of whole building to suggest the construction guideline for air tightness of low-energy building. Five out of six tested buildings showed 0.27 to 1.16 h^-1 of number of air changes except one community center. These buildings were carefully constructed not only for building planning but also for parts where there was a concern of air leakage, thereby securing high levels of air-tightness. The construction skills were developed as a checklist to manage and supervise the construction site. It is our suggestion to use this checklist at construction sites for ZEB with the high level of air-tightness.

• KIEAE Journal
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• 제13권3호
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• pp.79-89
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• 2013

The purpose of this study is to evaluate the airtightness performance of New Han-ok and to supply fundamental data for standards establishment. Air leakage testings were accomplished by means of blower door test in 26 bedrooms of 16 Han-oks located in Jeonnam happy villages. Followings are results. 1) Air change per hour at 50 Pa(ACH50) is located on 8.42~78.38. 2) No correlation between ACH50 and volumes, floor area, above grade surface area. 3) The more wood structural elements are exposed, attached spaces, wooden sliding and casement windows, the less airtightness performance. 4) An Airtightness with ACH50/20(NL, Normalized leakage) is located on 0.42~3.92 and building leakage class following F(4%), G(11%, sufficiently leaky, No need mechanical ventilation), H(4%, Need of cost-effective tightening), I(31%), J(50%) by a single-story house the normalized leakage of ASHRAE.

• 한국전기전자재료학회논문지
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• 제20권6호
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• pp.541-545
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• 2007

The critical current degradation behaviors of multifilamentary Bi-2223 superconducting tapes under pressurized liquid nitrogen were investigated using a r-shaped sample holder which gives a series of bending strains to tape. Three kinds of externally-reinforced Bi-2223 tapes with different hermeticity were used as samples. The tape with the thicker reinforcement layer had a better bending strain tolerance of $I_c$, but when the bending strain was calculated at the outermost filament, the $I_c$ degradation behavior became identical. For all samples, $I_{c0}$ decreased with the increase of applied pressure, but the $I_c$ degradation behavior with bending strain at each pressure level was similar. Furthermore, after depressurization from 1 MPa to atmospheric pressure, $I_c$ was completely recovered to its initial values. When the samples were warmed up to room temperature after pressurization tests, the ballooning damage occurred at lower bending strain regions. The region where ballooning was observed was identical to the one where the significant $I_c$ degradation occurred.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2006년도 하계학술대회 논문집 Vol.7
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• pp.257-258
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• 2006

The $I_c$ degradation behaviors of externally-reinforced Bi-2223 superconducting tapes under pressurized liquid nitrogen were investigated. Tapes with different thickness of reinforcement layers were compared and the results showed that when the bending strain was calculated at the outer surface, the tape with the thicker reinforcement showed a better bending strain tolerance of $I_c$, but when the bending strain was calculated at the outermost filament, the $I_c$ degradation behavior became identical for all tapes. $I_{c0}$ decreased with the increase of applied pressure. After depressurization to atmospheric pressure from 1 MPa, the $I_c$ was completely recovered to its initial values. Ballooning occurred after a thermal cycle.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• 설비공학논문집
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• 제25권9호
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• pp.493-498
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• 2013

The purpose of this study is to find the infiltration of detached houses, for energy consumption analyses. The pressurization and depressurization method is used to determine infiltration. Blower door tests are expressed in terms of ACH 50, which stands for the hourly air change rate at 50 Pa of fan pressure. The ACH 50 of existing Detached Houses ranges from 5 to 50. Air infiltration is related to construction year and accuracy, maintenance conditions, and so on. When estimating the infiltration of random detached houses, the year used can be the base value of the infiltration rate from 0.25 to 2.0 times/h. The maintenance conditions, construction accuracy and repair affect the air infiltration 0 to 0.5 times/h, 0 to 0.3 times/h, 0 to 1 times/h, respectively.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.