• Corrosion Science and Technology
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• 제5권3호
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• pp.94-98
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• 2006

The next generation nuclear power reactor will use Pb-Bi as the cooling material. The steel structure materials such as HT-9 used in the reactor suffer from corrosion when they are exposed to high temperature Pb-Bi. Therefore corrosion should be prevented to use Pb-Bi as the coolant material without any safety problem. One method is to control the oxygen content in Pb-Bi. An appropriate amount of oxygen in Pb-Bi can produce a thin oxide layer on steel, and this layer protects the steel from corrosion attack. Since the required oxygen content in Pb-Bi is in the range of $10^{-5}$ to $10^{-7}$ wt%, this small oxygen content can be controlled by flowing a mixture of hydrogen gas and water vapor. The stagnant corrosion test of HT-9 samples was performed by controlling the oxygen content up to 2,000 hours. The corrosion behavior of HT-9 was analyzed at the temperatures of $500^{\circ}C$ and $650^{\circ}C$ with a reduced condition and a oxygen content of $10^{-6}$ wt%.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1251-1260
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• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1999년도 학회창립 10주년 기념 1999년도 가을 학술발표회 논문집
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• pp.329-332
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• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as underground box structure, mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The objective of this paper is largely two folded. Firstly we introduce the cracks control technique by employing low-heat cement mix and thermal stress analysis. Secondly it show the application condition of the cracks control technique like sidewall of LNG in Inchonl.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1999년도 봄 학술발표회 논문집(I)
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• pp.293-298
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• 1999

The crack of concrete induced by the heat of hydration is a serious problem, particularly in concrete structures such as underground box structure, mat-slab of nuclear reactor buildings, dams or large footings, foundations of high rise buildings, etc.. As a result of the temperature rise and restriction condition of foundation, the thermal stress which may induce the cracks can occur. Therefore the various techniques of the thermal stress control in massive concrete have been widely used. One of them is prediction of the thermal stress, besides low-heat cement which mitigates the temperature rise, pre-cooling which lowers the initial temperature of fresh concrete with ice flake, pipe cooling which cools the temperature of concrete with flowing water, design change which considers steel bar reinforcement, operation control and so on. The objective of this paper is largely two folded. Firstly we introduce the cracks control technique by employing low-heat cement mix and thermal stress analysis. Secondly it show the application condition of the cracks control technique like the subway structure in Seoul.

• 한국유체기계학회 논문집
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• 제17권6호
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Korean Chemical Engineering Research
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• 제46권5호
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• pp.931-937
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• 2008

공정변수의 변화와 반응기의 성능을 정확하게 예측하기 위하여 Water Gas Shift Reaction(WGSR)을 위한 Multi-Tubular Reactor (MTR)의 상세 multiscale 모델링과 모사를 수행하였다. MTR은 비 균일 고체 촉매로 충진 된 4개의 관형반응기와 냉각을 위해 주변을 싸고 있는 shell side로 구성되어 있다. 유체의 흐름과 반응 kinetics가 반응기 성능에 큰 영향을 주고 있는 점을 고려할 때, Computational Fluid Dynamics (CFD)기법과 공정모델링 기법을 포함한 multiscale 방법론의 채택은 자연스럽고 필수 불가결한 일이다. $345^{\circ}C$로 관형반응기 부분으로 유입된 반응물은 반응의 결과 $390^{\circ}C$로 $45^{\circ}C$가량 온도가 증가하였으며, CO의 전환율은 0.89에 이르렀다. 쉘 사이드로 $190^{\circ}C$로 유입된 유체는 쉘 출구에서 $240^{\circ}C$로 약 $50^{\circ}C$ 가량의 온도 증가를 보였으며 이를 통하여 에너지 절감효과를 가져 올 수 있었으며 높은 전환율을 얻기 위해 반응기 부분의 온도를 적절히 제어할 수 있었다. 모사의 결과는 여러 문헌에 보고된 실험 결과와 매우 근접한 값을 나타내 본 연구를 통해 제시된 모델과 모사의 결과가 정확함을 알 수 있었다.