• Nuclear Engineering and Technology
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• 제56권1호
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• pp.257-264
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• 2024

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

• 에너지공학
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• 제24권3호
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• pp.135-141
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• 2015

중대사고시 핵연료와 원자로 내부 구조물이 용융되어 원자로용기의 하부에 재배치되면 밀도차이에 의하여 상부의 금속용융물층과 하부의 혼합물층으로 나누어진다. 하부 반구의 혼합물층에서는 지속적으로 붕괴열이 발생하고 이 열은 원자로용기의 건전성을 위협한다. 본 연구는 반구 내부의 체적 열원(Volumetric heat source)이 내재된 매질에서의 자연대류 열전달 현상을 물질전달 실험방법을 이용하여 모사하였다. 황산-황산구리의 구리도금계를 물질전달계로 사용하여 모사를 수행하였다. 수정 Rayleigh 수 $3{\times}10^{14}$에 대하여 Nusselt 수는 반구 하단에서 최소값을 보였고 곡면부를 따라 최상단으로 갈수록 증가하였다.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.964-974
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• 2020

The small modular reactors (SMRs) of the integrated pressurized water reactor (IPWR) type have been widely developed owing to their enhanced safety features. The SMR-IPWR adopts passive residual heat removal system (PRHRS) to extract residual heat from the core. Because the PRHRS removes the residual heat using the latent heat of the water stored in the emergency cooldown tank, the PRHRS gradually loses its cooling capacity after the stored water is depleted. A quick restoration of the power supply is expected infeasible under station blackout accident condition, so an advanced PRHRS is needed to ensure an extended grace period. In this study, an advanced design is proposed to indirectly incorporate a dry air cooling tower to the PRHRS through an intermediate loop called indefinite PRHRS. The feasibility of the indefinite PRHRS was assessed through a long-term transient simulation using the MARS-KS code. The indefinite PRHRS is expected to remove the residual heat without depleting the stored water. The effect of the environmental temperature on the indefinite PRHRS was confirmed by parametric analysis using comparative simulations with different environmental temperatures.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2088-2095
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• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2001년도 추계학술대회 논문집 전기물성,응용부문
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• pp.227-230
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• 2001

This paper divides the factors of an accident into two parts, that are electrical deterioration and thermal deterioration, to analyze a characteristic of the factor of an accident which can break out in the capacitor of metal vaporized polypropylene film. For the purpose of creating capacitor which is caused by electric deterioration, we applied DC overvoltage, induced self-healing and breakdown from element. We applied gradual heat to get an element which is cause by thermal deterioration. The chemical structure of the shape and surface is analyzed by thermogravimetric analyzer (TGA), Scanning Electron Microscope (SEM) and Fourier Transform Infrared Spectrometer(FT-IR). As a result, the peak of methylene group came out, in case of electrical deterioration, as observing the self-healing point. However, the peak is disappeared in the heat treated element by 500[$^{\circ}C$], and the peak of carbonyl group which has C=O came out in case of thermal deterioration.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.237-248
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• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• 한국화재소방학회논문지
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• 제23권2호
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• pp.1-5
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• 2009

본 연구에서는 분전반의 설치 형태를 조사하였고, 향 검지 시스템의 재현 실험을 통해서 전기 재해 예방의 가능성을 확인하고자 한다. 주택용 분전반 내에 설치된 차단기는 MCCB, RCD 순서로 되어 있으나, 산업용은 혼용되고 있다. 접속이 부적절하게 된 단자대에 가진기를 이용하여 진동을 가했을 때 불꽃이 확인되었다. 단자대에 향 캡슐을 부착하여 실험한 결과, 발생한 열에 의해 캡슐이 정확하게 작동되었다. 센서의 설치 위치에 따라 검지 시간의 차이는 약 10초 이었다. 향 검지 시스템은 전기 설비 사고 예방이 가능한 것으로 판단된다. 전선의 접속부에서 이상 발열이 있을 때 과열의 상태를 관리자에게 경보해 줌으로써 사고 예방이 가능할 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.650-659
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• 2016

As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

• 한국안전학회지
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• 제32권2호
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.

• Nuclear Engineering and Technology
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• 제17권1호
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• pp.8-15
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• 1985

Debris bed내·외로부터 깨스유량을 갖는 debris/water 열적상호작용 해석모델이 중대사고 분석을 위해 제시되었다. 제시된 모델은 증기 소비, debris bed에서 수소 생성, 유입깨스 및 화학반응열에 대한 인자들을 포함하고 있으며, 금속-물반응 및 debris/concrete 작용으로 인한 깨스 생성을 평가하기 위해 MARCH code에 도입시켰다. 그 결과 수소원은 격납용기 과도압력에 큰 영향을 미치나 debris bed로 대류깨스 냉각과 콘크리트로 전도 열손실은 debris bed 냉각성에 조그마한 영향을 주는 것으로 나타났다. 하지만 debris 인자의 재가열과 재용융은 콘크리트와 상호작용에 의해 상당히 지연될 수 있다.