• Nuclear Engineering and Technology
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• 제38권2호
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• pp.175-184
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• 2006

Boiling and flashing are driven by the same physics for nucleation, bubble growth, departure etc. An adequate model of boiling has to describe the flashing too. The subject of this paper is to prove this uniqueness of the elementary physics driving the both processes.

• Journal of Advanced Marine Engineering and Technology
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• 제38권5호
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• pp.581-588
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• 2014

Extensive researches toward pool boiling characteristics have been widely investigated. However, the correct understanding of its boiling crisis by Critical Heat Flux (CHF) phenomenon during steady and transient heat transfer as a fundamental database for designing heat generation systems is still need to be clarified. The pool boiling CHFs were investigated to clarify the generalized phenomena of transition to film boiling at transient condition. The CHFs were measured on 1.0 mm diameter horizontal cylinder of platinum for exponential heat generation rates with various periods for saturated liquids at atmospheric pressure. The incipience of boiling processes was completely different depending on pre-pressurization. Also, the dependence of pre-pressure in transient CHFs changed due to the wettability of boiling liquids. The trend of typical CHFs were clearly divided into the first, second and third groups for long, short and intermediate periods, respectively. By the effect of pre-pressurization, the boiling incipience mechanism was replaced from that by active cavities entraining vapor to that by the HSN in originally flooded cavies.

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.47-58
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• 2020

Most of the reactor trips in Korean NPPs related to core protection systems were caused not because of proximity of boiling crisis and, consequently, a damage in the core, but due to particular miscalculations or component failures related to the core protection system. The most common core protection system applied in Korean NPPs is the Core Protection Calculator System (CPCS), which is installed in OPR1000 and APR1400 plants. It generates a trip signal to scram the reactor in case of low Departure from Nucleate Boiling Ratio (DNBR) or high Local Power Density (LPD). However, is a reactor trip necessary to protect the core? Or could a fast power reduction be enough to recover the DNBR/LPD without a scram? In order to analyze the online calculation of DNBR/LPD, and the use of fast power reduction as trip avoidance methodology, a concept of CPCS with fast power reduction function was developed in Matlab® Simulink using systems engineering approach. The system was validated with maximum of 0.2% deviation from the reference and the dynamic deviation was maximum of 12.65% for DNBR and 6.72% for LPD during a transient of 16,000 seconds.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

• 한국초전도ㆍ저온공학회논문지
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• 제21권4호
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• pp.44-47
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• 2019

This paper is a study of recovery under load process of superconducting fault current limiter (SFCL). SFCL consists of five parallel-connected high-temperature superconductor (HTS) tapes additionally stabilized by stainless tape. Previously, HTS was heated by current pulse to simulate a short circuit in a power grid. During the cooling period, the current amplitude decreased to 23% or less of HTS critical current value, which is the simulation of network re-switching. When HTS with a polymer coating is cooled, temperature gradient on thermal insulation layer occurs, that prevents a boiling crisis and improves the heat sink into liquid nitrogen. Two samples are coated with a 30 ㎛ and 50 ㎛ polylactide (PLA) layers, reference sample has no polymer coating on it. Samples with a polymer coating show 3-5 times faster cooling than the reference one.

• 한국해양공학회지
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• 제22권1호
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• pp.13-21
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• 2008

In this study a method to predict CHF(Critical heat flux) in vertical round tubes with axially non-uniform cosine heat flux distribution for water was examined. For this purpose a local condition hypothesis based CHF prediction correlation for uniform heat flux in vertical round tubes for water was developed from 9,366 CHF data points. The local correlation consisted of 4 local condition variables: the system pressure(P), tube diameter(D), mass flux of water(G), and 'true mass quality' of vapor($X_t$). The CHF data points used were collected from 13 different published sources having the following operation ranges: 1.01 ${\leq}$ P (pressure) ${\leq}$ 206.79 bar, 9.92${\leq}$ G (mass flux) ${\leq}$ 18,619.39 $kg/m^2s$, 0.00102 ${\leq}$ D(diameter) ${\leq}$ 0.04468 m, 0.0254${\leq}$ L (length) ${\leq}$ 4.966 m, 0.11 ${\leq}$ qc (CHF) ${\leq}$ 21.41 $MVW/m^2$, and -0.87 ${\leq}X_c$ (exit qualities) ${\leq}$ 1.58. The result of this work showed that a uniform CHF correlation can be easily extended to predict CHF in axially non-uniform heat flux heater. In addition, the location of the CHF in axially non-uniform tube can also be determined. The local uniform correlation predicted CHF in tubes with axially cosine heat flux profile within the root mean square error of 12.42% and average error of 1.06% for 297 CHF data points collected from 5 different published sources.

• 식품과학과 산업
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• 제54권3호
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• pp.196-209
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• 2021

The origines of the Western roasting culture and East Asian boiling culture were studied and the importance of primitive pottery culture (8000-5000 BCE) in the Korea Strait coastal region was discussed. The primitive pottery culture probably initiated the Jjigae (stew) culture and the production of salt. It can be also postulated that fish fermentation, kimchi fermentation, and cereal alcohol fermentation originated during this period. Soybean culture emerged ca. 2,000 BCE in South Manchuria and the Korean Peninsula. This paper focuses on the role of Korean foodways in the food science and technology development for the sustainable and resilient food systems. We are facing a global food crisis caused by population growth, climate change, and high animal food consumption. Studies on the meat analog and cultured meat are the new trend in Food Science and Technology. The importance of the wisdom learned through the Northeast Asian traditional foods, for example, soybean curd (tofu) and meaty flavor production by fermentation for the research on the novel sustainable and resilient food systems are discussed.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.268-279
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• 2018

The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on $PWR-UO_2$ fuel rods with advanced claddings such as M5(R) under "low pressure-low temperature" or "high pressure-high temperature" water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on $UO_2$-M5(R) fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE-starting from base irradiation conditions it itself computes-is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5(R) is a trademark or a registered trademark of AREVA NP in the USA or other countries.