• Nuclear Engineering and Technology
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• 제41권3호
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.71-79
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• 2023

Air ingress incidents are major safety accidents in very-high-temperature reactors (VHTRs). Air containing a high volume fraction of oxygen may cause severe oxidation of core components at the VHTR, especially for the significantly thin alloy tube wall in the intermediate heat exchanger (IHE). The research objects of this study are Inconel 617 and Incoloy 800H, two candidate alloys for IHE in VHTR. The air ingress accident scenario is simulated with high-temperature air flow at 950 ℃. A continuous oxide scale was formed on the surfaces of both the alloys after the experiment. Because the oxide scale of Inconel 617 has a loose structure, whereas that of Incoloy 800H is denser, Inconel 617 exhibited significantly more severe internal oxidation than Incoloy 800H. Further, Inconel 617 showed a significant decrease in ultimate tensile strength and plasticity after aging for 200 h, whereas Incoloy 800H maintained its tensile properties satisfactorily. Through control experiment under vacuum, we preliminarily concluded that serious internal oxidation is the primary reason for the decline in the tensile properties of Inconel 617.

• 한국안전학회지
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• 제26권4호
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• pp.7-13
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• 2011

A study on reliability estimation of sequential-ordered multiple failure modes, which are sequentially ordered between failure modes in a considering system, was performed. Especially, an approach to estimate the probabilities of failure modes has been proposed under an assumption that failure modes are mutually exclusive and sequentially ordered by only a critical variable. A feasibility of the proposed approach were studied by a practical example, which is a reliability estimation of passive safety systems for a probabilistic safety assessment(PSA) of a very high temperature reactor(VHTR) that is under development as a future nuclear system with enhanced safety features. It is difficult to define a robust failure state of this nuclear system because of its enhanced radiation release characteristics, so the new approach is a useful concept to estimate not only its safety but also a PSA. A feasibility study applied two failure modes(e.g., small and large release of radioactive materials) with considering the integrated behavior of this nuclear system. It is expected that the multiple release states for a practical estimation can be easily extended to the aforementioned example. It was found out that the proposed approach was a useful technique to cover the unfavorable features of this nuclear system as to performing a VHTR PSA.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2008년도 추계학술대회 논문집
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• pp.63-64
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• 2008

• 대한기계학회논문집A
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• 제37권9호
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• pp.1189-1194
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• 2013

초고온가스로의 중간열교환기는 원자로에서 생산된 $950^{\circ}C$ 정도의 초고온 열을 수소생산 공장으로 전달하는 핵심 기기이다. 한국원자력연구원에서는 중간열교환기의 후보 형태로 고려되고 있는 인쇄기판형 열교환기의 실험실 수준 시제품을 제작하였다. 본 연구는 초고온헬륨루프 시험조건하에서 SUS316L 로 제작된 실험실 수준 인쇄기판형 열교환기 시제품의 고온구조건전성을 미리 평가하기 위한 작업의 일환으로 인쇄기판형 열교환기 실험실 수준 시제품에 대한 고온 구조해석 모델링, 거시적 열 해석 및 구조 해석을 수행하고 그 결과들을 정리한 것이다.

• 대한기계학회논문집A
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• 제35권9호
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• pp.1137-1143
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• 2011

초고온가스로의 중간열교환기는 원자로에서 생산된 $950^{circ}C$ 정도의 초고온 열을 수소생산 공장으로 전달하는 핵심 기기이다. 한국원자력연구원에서는 중간열교환기의 후보 형태로 고려되고 있는 인쇄기판형 열교환기의 소형 시제품을 제작하였다. 본 연구는 소형가스루프 시험조건하에서 인쇄기판형 열교환기 소형 시제품의 고온 구조건전성을 시험수행 전에 미리 평가하기 위한 작업의 일환으로 인쇄기판형 열교환기 소형 시제품에 대한 고온 구조해석 모델링, 거시적 열 해석 및 구조 해석을 수행하고 그 결과들을 정리한 것이다. 해석 결과는 곧 수행될 인쇄기판형 열교환기 소형 시제품 성능시험결과와 비교하고 또한 향후 제작될 중형 시제품 설계/제작에 반영할 것이다.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.99-108
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• 2006

SNU-RCCS is a water pool type RCCS (Reactor Cavity Cooling System) developed for VHTR (Very High Temperature Reactor) application by SNU (Seoul National University). Since radiation heat transfer is the major process of passive heat removal in a RCCS, it is important to determine the precise emissivity of the reactor vessel. Review studies have used a constant emissivity in the passive heat removal analysis, even though the emissivity depends on many factors such as temperature, surface roughness, oxidation level, wavelength, direction, atmosphere conditions, etc. Therefore, information on the emissivity of a given material in a real RCCS is essential in order to properly analyze the radiation heat transfer in a VHTR. The objectives of this study are to develop a method for compensation of the factors affecting the emissivity measurement using an infrared thermometer and to estimate the true emissivity from the measured emissivity via the developed method, especially in the SNU-RCCS environment. From this viewpoint, we investigated factors such as the attenuation effect of the window, filling gas, and the effect of background radiation on the emissivity measurements. The emissivity of the vessel surface of the SNU-RCCS facility was then measured using a sight tube. The background radiation was subsequently removed from the measured emissivity by solving a simultaneous equation. Finally, the calculated emissivity was compared with the measured emissivity in a separate emissivity measurement device, yielding good agreement with the emissivity increase with vessel temperature in a range of 0.82 to 0.88.

• 대한기계학회논문집B
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• 제34권5호
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• pp.553-561
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• 2010

한국원자력연구원에서는 초고온가스로를 모사할 수 있는 중형 헬륨 회로를 건설 중에 있다. 이 실험헬륨 회로에서 두 개의 전기 가열기가 헬륨 유체를 1 ~ 9 MPa 의 압력 하에서 $950^{\circ}C$ 까지 가열하게 된다. 이 실험 헬륨 회로의 설계 사양을 최적화하기 위하여, 본 연구에서는 두 개의 가열기 중 하류에 위치한 고온헬륨가열기 안의 복합열전달 현상을 전산유체역학으로 해석하였다. 해석 결과에서 헬륨 가열기 내 최대 온도는 허용 한계를 넘지 않았고, 이로써 선정된 기하구조의 열적 특성은 설계요건을 만족함이 확인되었다.

• 대한기계학회논문집A
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• 제34권10호
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• pp.1455-1462
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• 2010

초고온가스로에서 생성된 $950^{\circ}C$ 정도의 초고온 열을 이용하여 수소를 경제적이며 또한 대량으로 생산하기 위한 시스템이 원자력수소생산시스템이며, 이 시스템에서 공정열교환기는 초고온 열과 황-요오드 공정을 통해 수소를 생산하는 핵심 기기이다. 한국원자력연구원에서는 초고온가스로에 사용될 기기에 대한 성능시험을 위해 헬륨가스루프를 구축하고 공정열교환기 시제품을 제작하였다. 본 연구는 공정열교환기 시제품을 헬륨가스루프에서 시험하기 전에 미리 공정열교환기 시제품의 고온 구조건전성을 평가하기 위한 작업의 일환으로 공정열교환기 시제품에 대한 고온구조해석 모델링, 열해석 및 열팽창해석 결과들을 정리한 것이다. 해석 결과는 공정열교환기 시제품 성능시험 장치 설계에 반영할 것이다.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.1-6
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature gas cooled Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed only using parent material properties. In this study, high-temperature structural analysis using weld properties in weld zone was performed and the analysis results compared with the previous research.