• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4348-4358
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• 2022

The thermal-hydraulics phenomena in a containment during an accident will necessarily include radiative heat transfer (i) within the gas mixture due to the high radiative absorption and emission of steam and (ii) between the gas mixture and the surrounding structures. The analysis of some previous PANDA experiments (PSI, Switzerland) demonstrated the importance of the proper modelling of radiation for the benefit of numerical simulations. These results together with dedicated scoping calculations conducted for the present experiments indicated that the radiative heat transfer is considerable, even for a very low amount of steam (≈2%). The H2P2 series conducted in the large-scale PANDA facility at the Paul-Scherrer-Institut (PSI) in the framework of the OECD/NEA HYMERES-2 project is intended to enhance the understanding of thermal radiation phenomena and to provide a benchmark for corresponding numerical simulations. Thus, the test matrix was tailored around the two opposite extremes: either gas compositions with small steam content such that radiative heat transfer phenomena can be neglected. Or gas mixtures containing larger amounts of steam, so that radiative heat transfer is expected to play a dominant role. The H2P2 series consists of 5 experiments designed to isolate the radiation phenomena from convective and diffusive effects as much as possible. One vessel with a diameter of 4 m and a height of 8 m was preconditioned with different mixtures of air / steam at room and elevated temperatures. This was followed by the build-up of a stable helium stratification at constant pressure in the upper part of the vessel. After that, helium was injected from the top into the vessel which leads to an increase of the vessel pressure and a corresponding elevation-dependent and transient rise of the gas temperature. It is shown that even the addition of small amounts of steam in the initial gas atmosphere considerably impacts the radiative heat transport throughout all phases of the experiments and markedly influences i) the monitored gas peak temperature, ii) the temperature history during the compression and iii) the following relaxation phase after the compression was stopped. These PANDA experiments are the first of its kind conducted in a large scale thermal-hydraulic facility.

• 한국안전학회지
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• 제35권3호
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• pp.96-104
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• 2020

The numerical simulations were conducted to investigate the thermal-fluid phenomena occurred inside the experimental apparatus during a PCCS, used to remove heat released in accidents from a containment of light water nuclear power plant, operation. Numerical simulations of the flow and heat transfer caused by wall condensation inside the containment simulation vessel (CSV), which equipped with 18 vertical heat exchanger tubes, were conducted using the commercial computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the wall condensation model were used for turbulence closure and wall condensation, respectively. The simulation using the actual size of the apparatus. However, rather than simulating the whole experimental apparatus in consideration of the experimental cases, calculation resources, and calculation time, the simulation model was prepared only in CSV. Selective simulation was conducted to verify the effects of non-condensable gas(NC gas) concentration, CSV internal pressure, and wall sub-cooling conditions. First, as a result of the internal flow of CSV, it was observed that downward flow due to condensation occurred surface of the vertical tube and upward flow occurred in the distant place. Natural convection occurred actively around the heat exchanger tube. Due to this rising and falling internal flow, natural circulation occurred actively around the heat exchanger tubes. Next, in order to check the performance of built-in condensation model using according to the non-condensable gas concentration, CSV internal flow and wall sub-cooling, the heat flux values were compared with the experimental results. On average, the results were underestimated with and error of about 25%. In addition, the influence of CSV internal pressure and wall sub-cooling was small, but when the condensate was highly generated due to the low non-condensable gas concentration, the error was large compared to the experimental values. This is considered to be due to the nature of the condensation model of the CFX code. However, in spite of the limitations of CFD, it is valid to use the built-in condensation model of CFD for PCCS performance prediction from a conservative perspective.

• 전산구조공학
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• 제8권1호
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• pp.4-12
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• 1995

원자로 격납건물(Reactor Containment Bldg)은 정상가동시는 물론 냉각재상실사고(LOCA)를 포함하는 설계기준사고(DBA) 및 설계기준지진(DBE) 발생시 구조물 자체의 건전성 확보는 물론 주기기(NSSS Equipment)를 포함하는 안전관련 계통 및 기기를 안전하게 보호/지지하므로써 핵누출을 방지하여 발전소 종사자를 포함하는 국민의 재산과 생명을 보호하는 역할을 하는 원자력발전소에서 가장 중요한 구조물이다. 원자로 격납건물은 압력용기(Pressure Vessel : 설계내압 5 psi 이상인 용기)로 설계되는 격납용기와 1, 2차 차폐구조 등의 내부구조물로 구성되는데 이 중 본 소고에서는 격납용기의 해석 및 설계 그리고 구조건전성 시험 및 사용중검사에 대해서만 간략하게 기술한다.

• 대한기계학회논문집A
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• 제38권4호
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• pp.355-361
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• 2014

신규 원전에서 추진중인 외벽침수냉각 방식의 적용이 실패할 경우 노심용융물과 원자로공동 내유체의 상호작용으로 인해 증기폭발이 발생하며, 이는 격납건물 및 관통부 배관을 포함한 각 구조물의 건전성을 위협할 수 있다. 본 논문에서는 선행연구 분석결과를 토대로 증기폭발 현상을 모사할 수 있는 개선된 해석기법을 도출하고 알루미나 실험 모사를 통해 타당성을 확인하였다. 또한 동일한 기법을 원자로공동 해석에 적용하여 가상 파손위치에 따른 증기폭발 압력이력을 예측하였으며, 측면파손에 의한 최대압력 값이 하부파손에 의한 것보다 최대 70% 정도 높음을 보였다.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2311-2320
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• 2022

The main outcomes of the experiments H2P6 performed in the thermal-hydraulics large-scale PANDA facility at PSI in the frame of the OECD/NEA HYMERES-2 project are presented in this article. The experiments of the H2P6 series consists of two PANDA tests characterized by the activation of three (H2P6_1) or one (H2P6_2) cooler(s) in an initially stratified and pressurized containment atmosphere. The initial stratification is defined by a helium-rich region located in the upper part of the vessel and a steam/air atmosphere in the lower part. The activation of the cooler(s) results i) in the condensation of the steam in the vicinity of the cooler(s), ii) the corresponding activation of large scale natural circulation currents in the vessel atmosphere, with the result of iii) the re-distribution and mixing of the Helium stratification initially located in the upper half of the vessel and iv) the continuous pressure decay. The initial helium layer represents hydrogen generated in a postulated severe accident. The main question to be answered by the experiments is whether or not the interaction of the different, localized cooler units would be important for the application of numerical methods. The paper describes the initial and boundary conditions and the experimental results of the H2P6 series with the suggestion of simple scaling laws for both experiments in terms of i) the temperature difference(s) across the cooler(s), ii) the transient steam and helium content and iii) the pressure decay in the vessel. The outcomes of this scaling indicate that the interaction between separate, closely localized units does not play a prominent role for the present experiments. It is therefore reasonable to model several units as one large component with equivalent heat transfer area and total water flow rate.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3485-3490
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• 2007

The TEXAS-V code tuned for TROI-13 was used for analyzing the parametric findings in TROI experiments. The calculations on the melt composition are relatively similar to the TROI experimental results. The water depth effect in TEXAS-V code seems to be consistent with TROI experiments in some degree. The water area effect of TEXAS-V calculations seems not to be harmonious to that in TROI experiments. This seems to indicate that TEXAS-V as 1-dimensional code or as the numerical steam explosion has a limitation on estimating area effect. Thus, TEXAS-V tuned for TROI-13 seems to have an ability to estimate the parametric effect of TROI experiments. The evaluated TEXAS-V was used for estimating the ex-vessel steam explosion load. The calculated explosion pressure and load were about 40 MPa and 75 kPa.sec, which are not much threatening level for containment integrity, but are arguable value for the integrity.

• 대한기계학회논문집B
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• 제33권8호
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• pp.622-628
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• 2009

The TEXAS-V code tuned for TROI-13 was used for analyzing the parametric findings in TROI experiments. The calculations on the melt composition are relatively similar to the TROI experimental results. The water depth effect in TEXAS-V code seems to be consistent with TROI experiments in some degree. The water area effect of TEXAS-V calculations seems not to be harmonious to that in TROI experiments. This seems to indicate that TEXAS-V as 1-dimensional code or as the numerical steam explosion has a limitation on estimating area effect. Thus, TEXAS-V tuned for TROI-13 seems to have an ability to estimate the parametric effect of TROI experiments. The evaluated TEXAS-V was used for estimating the ex-vessel steam explosion load. The calculated explosion pressure and load were about 40 MPa and 75 kPa.sec, which are not much threatening level for containment integrity.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2836-2846
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• 2020

During severe nuclear power plant (NPP) accidents, a H₂/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H₂/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology for flammability evaluation of H₂/CO mixtures at ex-vessel phases characterized by three factors: CO concentration, high temperature, and diluents. The developed methodology adopted Le Chatelier's law and a calculated non-adiabatic flame temperature model. The methodology allows the consideration of the individual effect of the heat transfer characteristics of hydrogen and carbon monoxide on low flammability limit prediction. The accuracy of the developed model was verified using experimental data relevant to ex-vessel phase conditions. With the developed model, the prediction accuracy was improved substantially such that the maximum relative prediction error was approximately 25% while the existing methodology showed a 76% error. The developed methodology is expected to be applicable for flammability evaluation in chemical as well as NPP industries.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.