• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2702-2713
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• 2022

A steam generator tube rupture accompanying a core damage may cause the fission product to be released to environment bypassing the containment. In such an accident, the steam generator is the major path of the radioactive aerosol release. AEOLUS facility, the scaled-down model of Korean type steam generator, was built to examine the aerosol removal in the steam generator during the steam generator tube rupture accident. Integral and separate effect tests were performed with the facility for the dry and flooded conditions, and the decontamination factors were presented for different tube configurations and submergences. The dry test results were compared with the existing test results and with the analyses to investigate the aerosol retention physics by the tube bundle, with respect to the particle size and the bundle geometry. In the flooded tests, the effect of submergence were shown and the retention in the jet injection region were presented with respect to the Stokes number. The test results are planned to be used to constitute the aerosol retention model, specifically applicable for the analysis of the steam generator tube rupture accident in Korean nuclear power plants to evaluate realistic fission product behavior.

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1537-1546
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• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3164-3182
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• 2023

The CRUD on the fuel cladding under the pressurized water reactor (PWR) operating condition causes several issues. The CRUD can act as thermal resistance and increases the local cladding temperature which accelerate the corrosion process. The hideout of boron inside the CRUD results in axial offset anomaly and reduces the plant's shutdown margin. Recently, there are efforts to revise the acceptance criteria of emergency core cooling systems (ECCS), and additionally require the modeling of the thermal resistance effect of the CRUD during the performance analysis. There is an urgent need for the evaluation of the effect of the CRUD deposition on the cladding heat transfer under PWR operating conditions, but the experimental database is very limited. The experimental facility called DISNY was designed and constructed to analyze the CRUD-related multi-physical phenomena, and the performance analysis of the constructed DISNY facility was conducted. The thermal-hydraulic and water chemistry conditions to simulate the CRUD growth under PWR operating conditions were established. The design characteristics and feasibility of the DISNY facility were validated by the MARS-KS code analysis and separate performance tests. In the current study, detailed design features, design validation results, and future utilization plans of the proposed DISNY facility are presented.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.597-610
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• 2012

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+, which is intended to completely replace the conventional active auxiliary feedwater system. With an aim of validating the cooling and operational performance of PAFS, an experimental program is in progress at KAERI, which is composed of two kinds of tests; the separate effect test and the integral effect test. The separate effect test, PASCAL ($\underline{P}$AF$\underline{S}$ $\underline{C}$ondensing Heat Removal $\underline{A}$ssessment $\underline{L}$oop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in PAFS. A single, nearly-horizontal U-tube, whose dimensions are the same as the prototypic U-tube of the APR+ PAFS, is simulated in the PASCAL test. The PASCAL experimental result showed that the present design of PAFS satisfied the heat removal requirement for cooling down the reactor core during the anticipated accident transients. The integral effect test is in progress to confirm the operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. As the first integral effect test, an FLB (feedwater line break) accident was simulated for the APR+. From the integral effect test result, it could be concluded that the APR+ has the capability of coping with the hypothetical FLB accident by adopting PAFS and proper set-points of its operation.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1926-1938
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• 2020

We validated the performance of RELAP MOD3.3 code regarding the hybrid SIT with available experimental data. The concept of the hybrid SIT is to connect the pressurizer to SIT to utilize the water inside SIT in the case of SBO or SB-LOCA combined with TLOFW. We investigated how well RELAP5 code predicts the physical phenomena in terms of the equilibrium time, stratification, condensation against Separate Effect Test (SET) data. We also conducted the validation of RELAP5 code against Integrated Effect Test (IET) experimental data produced by the ATLAS facility. We followed conventional approach for code validation of IET data, which are pre-test and post-test calculation. RELAP5 code shows substantial difference with changing number of nodes. The increase of the number of nodes tends to reduce the condensation rate at the interface between liquid and vapor inside the hybrid SIT. The environmental heat loss also contributes to the large discrepancy between the simulation results of RELAP5 and the experimental data.

• 한국실내디자인학회논문집
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• 제19권5호
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• pp.29-37
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• 2010

The objective of correction is to let criminals receive education, be reformed and get vocational training during prison term by being separated from society and live as sound citizens after being released from prison. The facilities that execute such mission are referred to as corrective institution, and in Korea, some 47,000 persons are accommodated at 50 locations. Corrective facilities are facilities existing to separate and accommodate criminals safely, and return them to society after eliminating criminal nature from the convicts during the accommodation period. But corrective facilities in Korea representing the characteristics of corrective facilities vividly focusing on security. Therefore, while present pursuit of corrective facilities is to seek for reformation of criminals rather than separation of them from society, existing prison buildings are not appropriate facilitated at the moment. This study has the purpose to planning of indoor environment of corrective facility ward building as healing environment. Specifically, this study is contributing to propose a model of criminals healing environment equivalent to hospital facility to raise the effect of healing the criminal nature of criminals and cut the period of restoring them by applying healing environment elements and return the criminals to the society as successfully reformed social member.

• Korean Chemical Engineering Research
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• 제54권2호
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• pp.187-199
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• 2016

구미공단 대규모합성섬유제조업체들의 총괄 점오염원 및 개별 점오염원에 의한 각각의 트리메틸아민 관련 기여도 평가예측을 위하여 CALPUFF 모델링을 수행하였다. 또한 개별 점오염원의 총괄 점오염원에 대한 트리메틸 관련 상대적 기여도를 산출하였다. T사 배출설비 개선 후 총괄 점오염원의 트리메틸아민 배출에 의한 본 연구 CALPUFF 모델링 결과로서 2011년 겨울(1월), 봄(4월)에는 주로 구미 3공단에서, 여름(7월), 가을(10월)에는 주로 구미 1공단에서 트리메틸아민 농도의 최대값이 도출되었다. 한편 T사 배출설비 개선 후에는 구미 1공단의 경우에 트리메틸아민 예측 최대값 범위가 매우 개선되었음이 확인되었다. 반면에 구미3공단의 경우는 T사 배출설비 개선 후에 오히려 트리메틸아민 예측 최대값의 상한치가 더욱 커졌으나 그 상한치를 제외하였을 때는 T사 배출설비 개선 전의 트리메틸아민 측정치의 상한치와 비슷하였다. 이와 같이 T사 배출설비 개선이 공단별 트리메틸아민 농도에 미치는 영향이 구미 1공단에서는 컸으나 구미 3공단에서는 미미한 수준이었다. 한편 본 연구 CALPUFF 모델링 결과로서 트리메틸아민 농도 예측치는 트리메틸아민 농도의 측정치 범주와 거의 일치하므로 본 연구의 총괄 점오염원에 의한 트리메틸아민 기여도 평가에 대한 유효성을 보여주었다.

• 대한교통학회지
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• 제29권5호
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• pp.19-32
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• 2011

도로교통 안전분야에 투자되는 예산분야는 일반적으로 교통안전시설분야(Engineering), 교통단속분야(Enforcement), 교통안전교육 및 홍보분야(Education)의 3가지로 구분하고 있으며, 이것을 3E 정책이라 한다. 본 연구에서는 이러한 항목별 투자가 도로교통사고 감소에 어느 정도의 영향을 미치는 가에 대해 패널분석을 통해 추정하고자 하였다. 종속변수는 교통사고 건수, 사망자, 부상자수로 구분하였고, 설명변수로는 기존 연구에서 주로 이용되었던 인구, 자동차등록대수, 도로연장거리에 본 연구의 주된 관심사인 교통안전시설투자예산, 교통안전교육 및 홍보예산, 단속건수를 15개 지방자치단체의 지역자료와 1992년부터 2007년까지의 연도별 자료를 사용하였다. 전체 교통사고를 종속변수로 한 분석결과 동일한 투자액에 대비하여 교통안전교육 및 홍보투자가 시설투자에 비해 효과가 큰 것으로 분석되었다, 또한 6가지 세부교통사고로 분리한 분석결과에서도 동일한 결과를 보여 현재 교통 안전시설투자 위주로 이루어지고 있는 교통안전예산을 교통안전교육 및 홍보에도 분배될 필요성이 있음을 시사하고 있다.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.45-60
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• 2006

Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.

• Nuclear Engineering and Technology
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• 제33권3호
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• pp.315-326
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• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.