• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.127-132
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• 2001

An experimental study on penetration integrity of the reactor vessel has been performed under external vessel cooling during a core melt accident. In this study a series of experiments are performed for the verification of the effects of coolant in the annulus between the ICI(In-Core Instrumentation) nozzle and the thimble tube and also the effects of external vessel cooling on the integrity of the penetration using the test section including only one penetration and $Al_{2}O_{3}$ melt as a corium simulant. The experimental results have shown that penetration is more damaged in the case of no external vessel cooling compared with the case of external vessel cooling. It is preliminarily concluded that the external vessel cooling is very effective measure for the improvement of the penetration integrity. Also it is confirmed from the experimental results that the coolant in the annulus reduces the melt penetration distance through the annulus and enhance the integrity of the reactor vessel penetration in the end.

• Nuclear Engineering and Technology
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• 제47권4호
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• pp.443-453
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• 2015

A hybrid safety injection tank (H-SIT) can enhance the capability of an advanced power reactor plus (APR+) during a station black out (SBO) that is accompanied by a severe accident. It may a useful alternative to an electric motor. The operations strategy of the H-SIT has to be investigated to achieve maximum utilization of its function. In this study, the master logic diagram (i.e., an analysis for identifying the differences between an H-SIT and a safety injection pump) and an accident case classification were used to determine the parameters of the H-SIT operation. The conditions that require the use of an H-SIT were determined using a decision-making process. The proper timing for using an H-SIT was also analyzed by using the Multi-dimensional Analysis of Reactor Safety (MARS) 1.3 code (Korea Atomic Energy Research Institute, Daejeon, South Korea). The operation strategy analysis indicates that a H-SIT can mitigate five types of failure: (1) failure of the safety injection pump, (2) failure of the passive auxiliary feedwater system, (3) failure of the depressurization system, (4) failure of the shutdown cooling pump (SCP), and (5) failure of the recirculation system. The results of the MARS code demonstrate that the time allowed for recovery can be extended when using an H-SIT, compared with the same situation in which an H-SIT is not used. Based on the results, the use of an H-SIT is recommended, especially after the pilot-operated safety relief valve (POSRV) is opened.

• Nuclear Engineering and Technology
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• 제35권2호
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.

• Journal of Radiation Protection and Research
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• 제41권3호
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.124-133
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• 2017

In-vessel retention by passive external reactor vessel cooling under severe accident conditions is a viable approach for retention of radioactive core melt within the reactor vessel. In this study, a new and versatile coating technique known as "cold spray" that can readily be applied to operating and advanced reactors was developed to form a microporous coating on the outer surface of a simulated reactor lower head. Quenching experiments were performed under simulated in-vessel retention by passive external reactor vessel cooling conditions using test vessels with and without cold spray coatings. Quantitative measurements show that for all angular locations on the vessel outer surface, the local critical heat flux (CHF) values for the coated vessel were consistently higher than the corresponding CHF values for the bare vessel. However, it was also observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit vary appreciably along the outer surface of the test vessel. Nonetheless, results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward-facing boiling by > 88%.

• 대한기계학회논문집B
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• 제24권11호
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• pp.1447-1456
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• 2000

Simulated experiments of high pressure melt ejection(HPME) are performed to measure the released fraction of corium simulant from the French type PWR cavity. The experiments are carried out on a 1/20th linear scaled model of the Ulchin 1&2 cavity. Water or woods metal and nitrogen is used as simulant of molten corium and steam, respectively. Experimental parameters are water mass, annulus area and breach size. It is shown that only breach size effects is very important while the mass and the annulus area do not affect the released fraction. It is found that the liquid film transport is much more dominant mechanism than the entrainment droplet transport, especially in linear scale down simulated HPME experiment.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.397-400
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• 2002

Vapor explosion is one of the most important problems encountered in severe accident management of nuclear power plants. In spite of many efforts, a lot of questions still remain. So, KAERI launched a real experimental program called TROI using $UO_{2}$ and $ZrO_{2}$ to investigate the vapor explosion. Besides TROI tests, a small-scale experiment with molten-tin/water system was performed to quantify the characteristics of vapor explosion and to understand the phenomenology of vapor explosion. A vapor explosion was observed while the amount of air bubble and water temperature were systematically varied The mass and temperature of tin are $50\;g\;and\;150^{\circ}C$, respectively. Water temperature is set to $24^{\circ}C\;and\;50^{\circ}C$. The void fraction of air bubble ranges from $0\;to\;10\;{\%}$. The strength of vapor explosion was measured using dynamic pressure sensors attached in reactor tube wall. as a function of void fraction. In addition, a high speed video filming up to 1,000 flame/sec was taken in order to visually investigate the behavior of the vapor explosion .

• Nuclear Engineering and Technology
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• 제15권4호
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• pp.256-266
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• 1983

가압 경수로에서 극심한 사고시 Debris/Water/Concrete 상호작용에 의한 Debris Bed 냉각과 격납용기과도 압력 평가가 제시되었다. 이 논문에서 제시된 Debris/Water/Concrete 해석모델을 MARCH 전산코드에 도입시켜 TMLB'와 S$_2$D사고분류에 따라 현존 용융 모델과 비교할 때 저속의 콘크리트 분해율과 소량의 개스 생성을 나타내는 반면 입자형 모델은 물과 상호작용이 지배적이며, 더 높은 격납용기 압력을 야기시켰다. 그 결과 Debris Bed의 열전달에 미치는 개스 유입효과는 중요하지 않음이 입증되었다.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.907-914
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• 2015

The energetic steam explosion caused by contact between the high temperature molten core and water is one of the phenomena that may threaten the integrity of the containment vessel during severe accidents of light water reactors (LWRs). We examined the dependence of steam explosion loads in a typical reactor cavity geometry on selected model parameters and initial/boundary conditions by using a steam explosion simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA). Among the parameters, we put an emphasis on the water pool depth that has significance in terms of accident mitigation strategies including cavity flooding. The results showed a strong correlation between the load and the premixed mass, defined as the mass of the molten material in low void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flow rate were the primary factors to determine the premixed mass and the load. The water pool depth also showed a significant impact. The energy conversion ratio based on the enthalpy in the premixed mass was in a narrow range ~4%. Based on this observation, we proposed a simplified method for evaluation of the steam explosion load. The results showed fair agreement with JASMINE.