• Nuclear Engineering and Technology
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• 제41권5호
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.62-69
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• 2016

Current guidance considers that uniaxially loaded specimen with a deep crack is used for the determination of the ductile-to-brittle transition temperature. However, reactor pressure vessel is under biaxial loading in real and the existence of deep crack is not probable through periodic in-service-inspection. The elastic stress intensity factor and the elastic-plastic J-integral which were used for crack-tip stress field and fracture mechanics assessment parameters. The difference of the loading condition and crack geometry can significantly influence on these parameters. Thus, a constraint effect caused by differences between standard specimens and a real structure can over/underestimate the fracture toughness, and it affects the results of the structural integrity assessment, consequentially. The present paper investigates the constraint effects by evaluating the master curve $T_0$ reference temperature of PCVN (Pre-cracked Charpy V-Notch) and small scale cruciform specimens which was designed to simulate biaxial loading condition with shallow crack through the fracture toughness tests and 3-dimensional elastic-plastic finite element analyses. Based on the finite element analysis results, the fracture toughness values of a small scale cruciform specimen were estimated, and the geometry-dependent factors of the cruciform specimen considered in the present study were determined. Finally, the transferability of the test results of these specimens was discussed.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Nuclear Engineering and Technology
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• 제26권3호
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• pp.355-366
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• 1994

미국원자력규제위원회에서는 최근 안전해석에 최적전산코드의 사용을 허용하는 개정된 비상노심냉각계통 평가 규정을 제시하였다. 당 규정에서는 계통해석에 최적전산코드를 사용할 경우 불확실성 평가를 수행할 것을 요구하고 있다. 본 논문에서는 이러한 비상노심냉각계통의 규제요건을 만족하는 실제적인 최적평가방법론을 개발하여 대형냉각재상실사고에 적용하였다. 최적평가전산코드로는 RELAP5/MOD3.1을 개선한 RELAP5/MOD3/KAERI를 사용하였으며, 코드의 불확실성은 수개의 분리효과 및 총체효과 실험에 대한 평가를 수행함으로써 정량화 하였다. 적용대상 발전소로는 고리 3 & 4호기를 선정하였다. 민감도 분석을 통하여 응답방정식을 구성하였으며 각 응답방정식에 대하여 무작위 추출방식, Monte Carlo 방식으로 확률밀도함수를 구하였다. 최종 불확실성은 95%의 신뢰도로 정량화 하였으며 대형냉각재 상실사고시의 안전여유도에 대하여 논의하였다.

• 대한기계학회논문집A
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• 제37권5호
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• pp.665-671
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• 2013

제 4 세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX 와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX와 FHX는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다.

• Nuclear Engineering and Technology
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• 제9권4호
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• pp.223-236
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• 1977

가압수형 인자로에 사용되는 이산화우라 핵연료통의 역학적 열적설계 및 성능 분석을 위한 종합적 전산 코드가 개발되었다. PROD 1.0으로 명명된 이 코드에는 연료소자에서 반경 방향으로의 출력 침체, 연료소자의 균열, 고밀화 및 팽창, 핵분열기체의 방출, 피복관의 크립, 냉각수에 의한 열전달 및 부식층의 형성 둥의 제반 현상이 고려되었다. 이 FROD 1.0 코드로써 이차원적 온도 분포, 변형도, 응력 및 피복관 내압 등이 연소시간의 함수로서 적절한 전산 시간이내에 산출된다. 이 코드는 또한 종류가 다른 열중성자로에 쓰이는 산화 연료에도 응용필 수 있다. FROD 1.0의 응용으로서 원자로의 정상가동 상태와 미국 원자력학회 분류의 제 2상태에 해당하는 두 가지의 출력 경로에 더하여, 고리 원자력 발전소 1호기의 초기 노심에 장전된 핵연료봉의 연소특성을 예측하였다. 예측결과는 최종 안전 심사 보고서에 기술된 핵연료봉 설계기준과 비교되었으며 둘 사치의 차이점이 논의되었다.

• 한국압력기기공학회 논문집
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• 제7권4호
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• pp.12-18
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• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population. Many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solve the materials aging problem is integral to its success. A foundation for effective aging management of nuclear power plants is that aging is properly taken into account at each stage of a plant's lifetime, i.e. in design, manufacture, construction and operation including long term operation and decommissioning. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a regulation on the integrated materials aging management for nuclear facilities is proposed. The proposed regulation identifies key elements of effective aging management for nuclear power plants and provides the requirements on aging management for nuclear facilities throughout all stages of the lifetime of the plant.

• Nuclear Engineering and Technology
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• 제43권3호
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• 한국태양에너지학회 논문집
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• 제33권3호
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• pp.42-50
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• 2013

Solar thermal reactor was studied for hydrogen production with a two step thermochemical cycle including T-R(Thermal Reduction) step and W-D(Water Decomposition) step. NiFe2O4 and Fe3O4 supported by monoclinic ZrO2 were used as a catalyst device and Ni powder was used for decreasing the T-R step reaction temperature. Maintaining a temperature level of about $1100^{\circ}C$ and $1400^{\circ}C$, for 2-step thermochemical reaction, is important for obtaining maximum performance of hydrogen production. The controller was designed for adjusting high temperature solar thermal energy heating the foam-device coated with nickel- ferrite powder. A Pill temperature control system was designed based on 2-step thermochemical reaction experiment data(measured concentrated solar radiation and the temperature of foam device during experiment). The cycle repeated 5 times, ferrite conversion rate are 4.49~29.97% and hydrogen production rate is 0.19~1.54mmol/g-ferrite. A temperature controller was designed for increasing the number of reaction cycles related with the amount of produced hydrogen.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.