• Nuclear Engineering and Technology
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• 제54권3호
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• pp.849-859
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• 2022

Feed-water valves (FWVs) are used to regulate the flow rate of water entering steam generators, which are very important devices in nuclear power plants. Due to the working environment of relatively high pressure and temperature, there is strength failure problem of valve body in some cases. Based on the thermo-fluid-solid coupling model, the valve body stress of the feed-water valve in the opening process is investigated. The flow field characteristics inside the valve and temperature change of the valve body with time are studied. The stress analysis of the valve body is carried out considering mechanical stress and thermal stress comprehensively. The results show that the area with relatively high-velocity area moves gradually from the bottom of the cross section to the top of the cross section with the increase of the opening degree. The whole valve body reaches the same temperature of 250 ℃ at the time of 1894 s. The maximum stress of the valve body meets the design requirements by stress assessment. This work can be referred for the design of FWVs and other similar valves.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2207-2220
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• 2021

Lead alloy is used as coolant in Lead-based cooled Fast Reactor (LFR). The natural characteristics of lead alloy are combined with the simple structural design of LFR. This constitutes the inherent safety characteristics of LFR. The main work of this paper is to take the main coolant pump (MCP) in the lead-cooled fast reactor (LFR) as the research object, and to study the flow pattern distribution of the internal flow field under the reverse rotation pump condition, the reverse rotation positive-flow braking condition and the reverse rotation negative-flow braking condition. In this paper, the double-outlet volute type and the space guide vane are selected as the potential designs of the CLEAR-I MCP. In this paper, the CFD method is used to study the potential reverse accident of the MCP. It is found that the highest flow velocity in the impeller appears at the impeller outlet, and the Q-H curves of the two design programs basically coincide. The space guide vane type MCP has better hydraulic performance under the reverse rotation positive-flow condition, the Q-H curves of the two designs gradually separate with increasing flow rate, and the maximum flow velocity inside the space guide vane type MCP is obviously lower than that of the double-outlet volute type. For the reverse rotation test of MCP, only the condition of the forward rotating pump of the main coolant pump is tested and verified. For the simulation of the MCP in LBE medium, it proved that the turbulence model and basic settings selected in the simulation are reliable.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1556-1568
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.709-718
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• 2012

The pressurized water reactor APR1400 adopts DVI (Direct Vessel Injection) for the emergency cooling water in the upper downcomer annulus. The International Standard Problem number 50 (ISP-50) was launched with the aim to investigate thermal hydraulic phenomena during a 50% DVI line break scenario with best estimate codes making use of the experimental data available from the ATLAS facility located at KAERI. The present work describes the calculation results obtained for the ISP-50 using the RELAP5/MOD3.3 system code. The work aims at validation and assessment of the code to reproduce the observed phenomena and investigate about its limitations to predict complicated mixing phenomena between the subcooled emergency cooling water and the two-phase flow in the downcomer. The obtained results show that the overall trends of the main test variables are well reproduced by the calculations. In particular, the pressure in the primary system show excellent agreement with the experiment. The loop seal clearance phenomenon was observed in the calculation and it was found to have an important influence on the transient progression. Moreover, the collapsed water levels in the core are accurately reproduced in the simulations. However, the drop in the downcomer level before the activation of the DVI from safety injection tanks was underestimated due to multi-dimensional phenomena in the downcomer that are not properly captured by one-dimensional simulations.

• 한국태양에너지학회 논문집
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• 제31권1호
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• pp.15-22
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• 2011

It is a great interest to convert more energy in the heat source into the power and to improve the efficiency of power generating processes. Since the efficiency of power generating processes becomes poorer as the temperature of the source decreases, to use an ammonia-water mixture instead of water as working fluid is a possible way to improve the efficiency of the system. In this work performance of ammonia-water regenerative Rankine cycle is investigated for the purpose of extracting maximum power from low-temperature waste heat in the form of sensible energy. Special attention is paid to the effect of system parameters such as mass fraction of ammonia and turbine inlet pressure on the characteristics of system. Results show that the power output increases with the mass fraction of ammonia in the mixture, however workable range of the mass fraction becomes narrower as turbine inlet pressure increases and is able to reach 16.5kW per unit mass flow rate of source air at $180^{\circ}C$.

• Nuclear Engineering and Technology
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• 제53권4호
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• pp.1100-1108
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• 2021

Supercritical water has great potential as a coolant for nuclear reactor. Its use will lead to higher thermal efficiency of Rankine cycle. However, in certain conditions heat transfer may get deteriorated which may lead to undesirable high clad surface temperature. It is necessary to estimate the operating conditions in which heat transfer deterioration (HTD) will take place, so as to establish thermal margins for safe reactor operation. In the present work, the heat flux corresponding to onset of HTD for vertically upward flow of supercritical water in a pipe is obtained over a wide range of system parameters, namely pressure, mass flux, and pipe diameter. This is done by performing large number of simulations using an in-house CFD code, which is especially developed and validated for this purpose. The identification of HTD is based on observance of one or more peak/s in the computed wall temperature profile. The existing correlations for predicting the onset of HTD are compared against the results obtained by present simulations as well as available sets of experimental data. It is found that the prediction accuracy of the correlation proposed by Dongliang et al. is best among the existing correlations.

• 한국수자원학회논문집
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• 제40권4호
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• pp.325-334
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• 2007

본 연구에서는 점진적인 유량 및 압력이 변화하는 상수관망에서 Rigid Water Column Theory를 이용하여 정상모형의 확장기간 모의해석보다 정확하고 수충격 해석보다는 계산비용 및 노력 측면에서 효율적으로 장시간 부정류 해석 모형을 개발하였다. 개발된 모형을 이용하여 실제관망에 대하여 24 시간 열 수요량을 고려한 부정류 해석 및 밸브폐쇄로 인한 수충격해석 모의에 적용하였고 해석 결과는 다음과 같다. 24 시간 일변화 모의의 경우에 수요량이 증가할 경우 모든 관로에서 압력감소가 나타났으며 수요량이 감소할 경우 압력증가가 나타났다. 그리고 일 수요량의 변화에 따라 나타난 절점에서의 유량 및 압력 변화폭은 각 절점마다 다르고 수요량과 유량의 변화양상이 반대로 나타나는 관로도 발생하고 있으며 KYPIPE2의 결과와 본 모형의 유량 및 압력차이도 발생하고 있어 상수관망의 동역학적 해석의 필요성이 대두되었다. 밸브폐쇄로 인한 수충격모의에 본 모형이 적용되었을 때 본 모형은 유체의 압축성을 무시함으로 인해 밸브 완전 폐쇄와 동시에 압력과 유량의 변화가 전 관망에 발생하였고 수충격모형은 유체의 탄성으로 인해 발생된 압력파의 도달시간이 필요함으로 압력과 유량변화가 지체되어 나타났으나 전체적인 변화양상 및 변화폭의 크기 등은 유사한 경향을 나타내어 본 모형의 적용성을 입증하였다. 본 연구에서 개발된 프로그램은 장기간 점진적인 관로 부정류를 비교적 정확하게 해석할 수 있을 것으로 판단되며 이를 이용하여 관로내 오염물의 확산해석, 수요량을 고려한 절점에서의 압력제어 및 누수저감, 장기간 관로내의 유량 및 압력 변화를 고려한 관망관리 등의 분야에서 효율적으로 이용될 수 있을 것으로 기대되었다.

• Journal of Information Technology Applications and Management
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• 제25권4호
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• pp.185-196
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• 2018

This study investigates the social network among authors to improve the quality of Panel researches. Korea Health Panel (KHP), implemented by the collaborative work between KIHASA (Korea Institute for Health and Social Affairs) and NHIC (National Health Insurance Service) since 2008, provides a critical infrastructure for policy making and management for insurance system and healthcare service. Using bibliographic data extracted from academic databases, eighty articles were extracted in domestic and international journals from 2008 to 2014, April. Data were analyzed by NetMiner 4.0, social network analysis software, to identify the extent to which authors are involved in healthcare use research and the patterns of collaboration between them. Analysis reveals that most authors publish a very small number of articles and collaborate within tightly knit circles. Centrality measures confirm these findings by revealing that only a small percentage of the authors are structurally dominant, and influence the flow of communication among others. It leads to the discovery of dependencies between the elements of the co-author network such as affiliates in health panel communities. Based on these findings, we recommend that Korea Health Panel could benefit from cultivating a wider base of influential authors and promoting broader collaborations.