• 대한기계학회논문집B
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• 제23권7호
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• pp.831-840
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• 1999

In this work, an aerothermodynamic calculation model for cooled axial flow turbine blades with trailing edge ejection is suggested and a mean line performance analysis of a turbine stage with nozzle cooling is carried out. A unique model regarding the interaction between coolant and main gas is proposed, while existing correlations are adopted to predict viscous loss and blade outflow angle. The interactions considered are the heat transfer from main gas to coolant and the temperature and pressure losses by the mixing of two streams due to the trailing edge coolant ejection. For a stator blade without ejection, trailing edge loss calculated by the trailing edge analysis is compared with that calculated by loss correlation. The effect of heat transfer effectiveness of coolant passage on the mixing loss is analyzed. For a model turbine stage with nozzle cooling, parametric analyses are carried out to investigate the effect of main design variables(coolant mass flow ratio, temperature and ejection area) on the stage performance.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.179-189
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• 1994

이상유동시 원자로 냉각재 펌프의 성능을 펌프의 기하학적 형상 및 단상 유동시의 펌프 성능을 이용하여 예측하였다. 단상 유동시의 원자로 냉각재 펌프의 벽면 마찰손실은 Truckenbrodt의 경계층 이론을 이용하여 예측하였으며, 계산된 벽면 마찰 손실 및 분리 손실을 사용하여 이상유동시의 수두손실을 예측하였다. 해석결과는 Combustion Engineering 사의 펌프 실험 데이터와 비교하였다. 또한 냉각재 상실사고시 이상유동배수가 첨두 피복재 온도에 미치는 영향을 RELAP5를 사용하여 평가하였으며, 분석결과는 이상유동배수의 정확성이 중요한 영향을 미치는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1532-1539
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• 2019

This paper analyzes a new method for reducing boiling water reactor fuel temperature during a Loss of Coolant Accident (LOCA). The method uses a device called Reverse Flow Restriction Device (RFRD) at the inlet of fuel bundles in the core to prevent coolant loss from the bundle inlet due to the reverse flow after a large break in the recirculation loop. The device allows for flow in the forward direction which occurs during normal operation, while after the break, the RFRD device changes its status to prevent reverse flow. In this paper, a detailed simulation of LOCA has been carried out using the U.S. NRC's TRACE code to investigate the effect of RFRD on the flow rate as well as peak clad temperature of BWR fuel bundles during three different LOCA scenarios: small break LOCA (25% LOCA), large break LOCA (100% LOCA), and double-ended guillotine break (200% LOCA). The results demonstrated that the device could substantially block flow reversal in fuel bundles during LOCA, allowing for coolant to remain in the core during the coolant blowdown phase. The device can retain additional cooling water after activating the emergency systems, which maintains the peak clad temperature at lower levels. Moreover, the RFRD achieved the reflood phase (when the saturation temperature of the clad is restored) earlier than without the RFRD.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.474-483
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• 2021

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters-burst stress, burst strain, burst temperature-during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.812-817
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• 1998

Experimental studies on heat transfer and solidification of the molten metal pool with overlying coolant with boiling were performed The simulant molten pool material is tin (Sn) with the melting temperature of 232$^{\circ}C$. Demineralized water and R113 are used as the working coolant. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The Nusselt number and the Rayleigh number in the molten metal Pool region of this study are compared between the water coolant case and the R113 coolant case. The experimental results or the water coolant are higher than those for R113. Also, the empirical relationship of the Nusselt number and the Rayleigh number is compared with the literature correlations measure from mercury. The present experimental results are higher than the literature correlations. It is believed that this discrepancy is caused by the effect of heat loss to the environment on the natural convection heat transfer in the molten pool.

• 에너지공학
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• 제26권3호
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• pp.71-77
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• 2017

피동고압충수용 혼합형 안전주입탱크 (Hybrid SIT)의 압력평형시간은 냉각수 주입시기를 결정하는 주요인자이다. 한국원자력연구원 (KAERI)에서는 Hybrid SIT에서의 내부 열수력적 거동을 고찰하기 위해 개별효과시험 장치를 구축하였으며, 다양한 운전조건에서의 압력평형시간에 대한 민감도 시험을 수행하였다. 개별효과시험을 통해 압력평형시간을 결정하는 주요인자들을 도출하였으며, 그 중 증기의 벽면응축 및 냉각재와의 직접접촉응축이 압력평형시간을 결정하는 주요 현상임을 파악하였다. 본 연구에서는 개별효과 시험결과들을 이용하여 각각의 응축현상들이 압력평형에 미치는 영향을 정량적으로 분석하고 혼합형 SIT의 압력평형시간을 예측하기 위한 방법론을 제시하였다.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.979-988
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• 2017

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.