• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.19-31
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• 1994

The objective of this study is to conduct a thermal-hydraulic analysis on the spent fuel pool and to evaluate a parametric effect for the thermal-hydraulic analysis of spent fuel pool. The selected parameters are the Reynolds Number and the gap flow through the oater gap between fuel cell and fuel bundle. The simplified flow network for a path of fuel cells is used to analyze the natural circulation phenomenon. In the flow network analysis, the pressure drop for each assembly from the entrance of the fuel rack to the exit of the fuel assembly is balanced by the driving head due to the density difference between the pool fluid and the average fluid in each spent fuel assembly. The governing equations ore developed using this relation. But, since the parameters(flow rate, pressure loss coefficient, decay heat, density)are coupled each other, iteration method is used to obtain the solution. For the analysis of the YGN 3&4 spent fuel rack, 12 channels are considered and the inputs such as decay heat and pressure loss coefficient are determined conservatively. The results show the thermal-hydraulic characteristics(void fraction, density, boiling height)of the YGN 3&4 spent fuel rack. There occurs small amount of boiling in the cells. Fuel cladding temperature is lower than 343.3$^{\circ}C$. The evaluation of parametric effect indicates that flow resistances by geometric effect are very sensitive to Reynolds number in the transition region and the gap flow is negligible because of the larger flow resistance in the gap flow path than in the fuel bundle.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.405-418
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• 2019

Based on spent fuels characteristics from domestic nuclear power plants and a disposal scenario from the current basic plan for high-level radioactive waste management, an improved disposal system has been proposed that enhances disposal efficiency and economic effectiveness compared to the existing disposal system. For this purpose, two disposal canisters concepts were derived from the length of the spent fuel generated from the nuclear power plants. In the disposal scenario, the acceptable amount of decay heat for each disposal container was determined, taking into account the discharge and disposal times of spent fuels in accordance with the current basic plan. Based on the determined decay heat of the two types of disposal canisters and the associated disposal system, thermal stability analyses were performed to confirm their suitability to the proposed disposal system design requirement and disposal efficiency assessment. The results of this study confirm 20% reduction in the disposal area and 20% increase in disposal density for the proposed disposal system compared to the existing system. These results can be used to establish a spent fuel management policy and to design a viable commercial disposal system.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1396-1408
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• 2020

Filtered containment venting system (FCVS) is one of the severe accident mitigation systems designed to release containment pressurization to maintain its integrity. The thermal-hydraulic behaviors in FCVSs are important since they affect the operation characteristics of the FCVS. In this study, a representative FCVS was modeled by RELAP5/Mod3.3 code, and the Station BlackOut (SBO) was chosen as an accident scenario. The thermal-hydraulic behaviors of an FCVS during long-term operation with two venting strategies (open-and-close strategy, open-and-non-close strategy) and the sensitivity analysis of important parameters were investigated. The results show that the FCVS can operate up to 250 h with a periodic open-and-close strategy during an SBO. Under the combined effects of steam condensation and water evaporation, the solution inventory in the FCVS increases during the venting phase and decreases during the intermission phase, showing a periodic pattern. Under this condition, the appropriate initial water level is 3-4 m; however, it should be adjusted according to the environment temperature. The FCVS can accommodate a decay heat power of 150-260 kW and may need to feed water for a higher decay heat power or drain water for a lower decay heat power during the late phase. The FCVS can function within an opening pressure range from 450 kPa to 500 kPa and a closing pressure range between 250 kPa and 350 kPa. When the open-and-non-close strategy is adopted, the solution inventory increases quickly in the early venting phase due to steam condensation and then decreases gradually due to the evaporation of water; drying-up may occur in the late venting phase. Decreasing the venting pipe diameter and increasing the initial water level can mitigate the evaporation of the scrubbing solution. These results are expected to provide useful references for the design and engineering application of FCVSs.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.395-409
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• 2000

Parametric studies were performed to assess the sensitivity in determining the maximum in-vessel heat removal capability from the core material relocated into the lower plenum of the reactor pressure vessel (RPV）during a core melt accident. A fraction of the sensible heat can be removed during the molten jet delivery from the core to the lower plenum, while the remaining sensible heat and the decay heat can be transported by rather complex mechanisms of the counter-current flow limitation (CCFL) and the critical heat flux (CHF）through the irregular, hemispherical gap that may be formed between the freezing oxidic debris and the overheated metallic RPV wall. It is shown that under the pressurized condition of 10MPa with the sensible heat loss being 50% for the reactors considered in this study, i.e. TMI-2, KORI-2 like, YGN-3&4 like and KNGR like reactors, the heat removal through the gap cooling mechanism was capable of ensuring the RPV integrity as much as 30% to 40% of the total core mass was relocated to the lower plenum. The sensitivity analysis indicated that the cooling rate of debris coupled with the sensible heat loss was a significant factor The newly proposed heat removal capability map (HRCM) clearly displays the critical factors in estimating the maximum heat removal from the debris in the lower plenum. This map can be used as a first-principle engineering tool to assess the RPV thermal integrity during a core melt accident. The predictive model also provided ith a reasonable explanation for the non-failure of the test vessel in the LAVA experiments performed at the Korea Atomic Energy Research Institute (KAERI), which apparently indicated a cooling effect of water ingression through the debris-to-vessel gap and the intra-debris pores and crevices.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.11
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• pp.991-998
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• 2010

The effect of increasing the elevation of an IHX (intermediate heat exchanger) on the transient performance of the KALIMER-600 reactor pool during the early phase of a loss of normal heat sink accident was investigated. Three reactors equipped with IHXs that were elevated to different heights were designed, and the thermal-hydraulic analyses were carried out for the steady and transient state by using the COMMIX-1AR/P code. In order to analyze the effects of the elevation of an IHX between reactors, various thermal-hydraulic properties such as mass flow rate, core peak temperature, RmfQ (ratio of mass flow over Q) and initiation time of decay heat removal via DHX (decay heat exchanger) were evaluated. It was found that with an increase in the IHX elevation, the circulation flow rate increases and a steep rise in the core peak temperature under the same coastdown flow condition is prevented without a delay in the initiation of the second stage of cooling. The available coastdown flow range in the reactor could be increased by increasing the elevation of the IHX.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• Tunnel and Underground Space
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• v.8 no.1
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• pp.26-36
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• 1998

To assess the groundwater flow near high-level radioactive waste repositories, it is important to understand the effect of coupling among thermal, hydraulic, and mechanical effects. In this paper, detailed mathematical approach to model the groundwater flow near the waste form surrounded by buffer, influenced by decay heat of radioactive waste along with stress change is developed. Two cases(1) before the full expansion of buffer and (2) after the full expansion of buffer are modelled. Based on the mathematical models in this paper, detailed numerical study shall be pursued later.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.201-206
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• 1996

건식가공(Dry Process)이 사용전,후 DUPIC 핵연료의 붕괴열(Decay Heat), Hazard Index, 조사선량률(Dose Rate) 등에 미치는 영향을 계산하고, 그 원인을 분석하였다. DUPIC 사용방안으로 표준 연소도(35,000 MWD/MTU)의 경우와 장주기 연소도(50,000 MWD/MTU)의 경우를 고려하여 계산하였으며, DUPIC핵연료는 20년 냉각후 가공하는 것을 기준으로 하였다. 또한 DUPIC핵연료 장전시 고려할 수 있도록 사용전 DUPIC 핵연료에 대한 계산을 핵연료 집합체(Bundle) 단위로 하였다. 조사선량과 붕괴열은 건식가공에 상당히 민감한 반응을 보였고 이는 주로 Cs의 제기에 의한 것이다.

• Proceedings of the Korean Vacuum Society Conference
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• 2013.08a
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• pp.320.1-320.1
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• 2013

We have investigated the effect of surface passivation for crystalline silicon solar cell using ozone-based atomic layer deposited (ALD) $Al_2O_3$. We examined passivation properties such as uniformity, carrier lifetime, thickness, negative fixed charge density at AlOx/Si interface, and reflectance. The influences of process temperature and heat treatment were investigated using microwave photoconductance decay (PCD). Ozone-based ALD $Al_2O_3$ film shows the best carrier lifetime at lower deposition temperature than $H_2O$-based ALD.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.759-766
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• 2013

As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III) nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+), and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX). To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL). The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.