• Nuclear Engineering and Technology
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• 제45권7호
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• 한국정보통신학회:학술대회논문집
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• 한국해양정보통신학회 1999년도 춘계종합학술대회
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• pp.394-397
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• 1999

임팩트이온화는 고전계하에서 고에너지를 지닌 캐리어간 산란으로써 전자전송해석에 필수적인 요소이다. 그러나 포물선 또는 비포물선 E-k관계는 고에너지에서 실제 밴드구조와 매우 상이한 결과를 나타내므로 임팩트이온화에 대한 정화한 모델은 풀밴드 E-k관계와 페르미의 황금법칙을 이용하여 제시하고 있다. 본 연구에서는 풀밴드를 이용하여 실리콘 임팩트이온화율에 영향을 미치는 상태밀도와 에너지밴드구조 관계, overlap integral의 에너지에 대한 변화를 조사 분석하였다. 실리콘의 에너지밴드구조를 구하기 위하여 경험적 의사포텐셜방법을 사용하였으며 상태밀도를 구하기 위하여 사면체방법을 사용하였다. 또한 임팩트이온화율을 구하기 위하여 페르미의 황금법칙들 사용하였다. 결과적으로 상태밀도는 에너지증가에 따라 단조 증가하는 임팩트이온화율과 달리 에너지밴드구조에 따라 변화하였다. 그러나 overlap integral은 에너지증가에 따라 큰 값을 갖는 분포가 증가함으로써 임팩트이온화율에 직접 영향을 미치는 것을 알 수 있었다.

• 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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• 제38권2호
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• pp.185-194
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• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• Nuclear Engineering and Technology
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• 제33권1호
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• pp.111-120
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• 2001

In this study, a dynamic analysis of the integral reactor SMART (System-integrated Modular Advanced ReacTor) under postulated seismic events is performed to review the response characteristics of the major components. To enhance the feasibility of an analysis model, a detailed finite element model is synchronized with the products of concurrent design activities. The artificial time history, which has been applied to the seismic analysis for the Korean Standard Nuclear Power Plant (KSNP), is chosen to envelop broad site specifics in Korea. Responses in the horizontal direction are found slightly amplified, while those in the vertical direction are suppressed. Since amplified response is monitored at the control element drive mechanism (CEDM), minor design provision is considered to enhance the integrity of the subsystem.

• Nuclear Engineering and Technology
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• 제40권3호
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• pp.199-212
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• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• 천문학회지
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• 제29권spc1호
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• pp.315-316
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• 1996

Based on X-ray (1-8 ${\AA}$) flux data for 1972-1995 the integral spectra of solar flare energy were computed. It has been shown that the spectral index $\beta$ of the integral energy spectrum (IES) vanes systematically with the 11-year cycle phase. The interval of $\beta$-variations (0.47 <$\beta$<1) is characteristic of UV-Cet stars. The maximum energy of the X-ray flares does not exceed $10^{32}$ erg.

• 대한기계학회논문집
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• 제19권1호
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• pp.170-180
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• 1995

A parallel crack in bonded dissimilar orthotropic planes under out-of-plane loading is analyzed. The problem is formulated by Fourier integral transforms, and reduced to a pair of dual integral equations. By solving the integral equations, the asymptotic stress and displacement fields near the crack tip are determined in closed form, from which the stress intensity factor and energy release rate are obtained. Discontinuity in the stress intensity factor as the distance ratio h/a of the parallel crack approaches zero is found, while the energy releas rate is shown to be continuous at h/a = 0. This information can immediately be used to generate the stress intensity factor for the parallel crack near the interface. By employing "the maximum energy release rate criterion", it could be shown in the case of no existing crack initially that the parallel crack is formed far from the interface for the more compliant material, while it is formed close to the interface for the stiffer material. material.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3551-3566
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• 2022

The STELLA program was launched to support the PGSFR development in 2012 and for the 2nd stage, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS, IHTS and DHRS. In STELLA-2, the long-term transient behavior after accidents can be observed and the overall safety aspect can also be evaluated. In this paper, the basic design concept from engineering basis to specific design is described. The design was aimed to meet similarity criteria and requirements based on various non-dimensional numbers and the result satisfied the key features to explain the reasoning of safety evaluation. The result of this study was used to construct the facility and the experiment is on-going. In general, the final design meets the similarity criteria of the multidimensional physics inside the reactor pool. And also, for the conservation of natural circulation phenomena, the design meets the similarity requirements of geometry and thermo-dynamic behavior.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.826-833
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• 2022

A two-phase natural circulation test using SMART integral test loop (SMART-ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler-condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART-ITL was compared with those in pressurized water reactor simulators. In the SMART-ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.