• Nuclear Engineering and Technology
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• v.54 no.6
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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.

• Journal of the Korean Institute of Gas
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• v.4 no.1 s.9
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• pp.69-76
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• 2000

Explosion Quantities and flashing mass resulting from the variation of temperature are calculated by a computer program, BLEVE ESTIMATOR, to carry out the risk assessment of BLEVE. The damages caused by the BLEVE are estimated under the explosion of the simulation condition similar to the Puchun LP gas station accident, and the results are compared with the commercial program SAFER of Dupont CO. Explosion quantities and flashing mass increase exponentially with the increase of explosion temperature. These values for propane are relatively higher than those for n-butane. In conditions of higher vessel temperature, vessel pressure, and liquid ratio of containment, higher overpressures are calculated.

• Structural Engineering and Mechanics
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• v.60 no.2
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• pp.337-350
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• 2016

Even though extensive researches have been performed for steam explosion due to their complex mechanisms and inherent uncertainties, establishment of severe accident management guidelines and strategies is one of state-of-the arts in nuclear industry. The goal of this research is primarily to examine effects of vessel failure modes and locations on nuclear facilities under typical steam explosion conditions. Both discrete and integrated models were employed from the viewpoint of structural integrity assessment of steel components and evaluation of the cracking and crushing in reinforced concrete structures. Thereafter, comparison of systematic analysis results was performed; despite the vessel failure modes were dominant, resulting maximum stresses at the all steel components were sufficiently lower than the corresponding yield strengths. Two failure criteria for the reinforced concrete structures such as the limiting failure ratio of concrete and the limiting strains for rebar and liner plate were satisfied under steam explosion conditions. Moreover, stresses of steel components and reinforced concrete structures were reduced with maximum difference of 12% when the integrated model was adopted comparing to those of discrete models.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4003-4013
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• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2008.04a
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• pp.166-171
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• 2008

LNG carrier is a special-purpose vessel to transport natural gas (NG) from the place of origin to each consuming country. To increase the capacity of canying LNG carrier, the natural gas is conveyed as a state of liquid called LNG (Liquefied Natural Gas) during a voyage because the total volume of NG is surprisingly reduced when it is cooled down to $-162^{\circ}C$. That is why the design of insulation of the carriers is important to protect LNG from the external heat invasion, and it has been a great challenging subject for several decades in the shipbuilding industry. For this ultimate goal, the boil-off rate (BOR) needs to be accurately estimated during a voyage. Therefore, the goal of this study is to propose a numerical method for estimating the BOR of LNG for given insulation containment subject to external temperature conditions during voyage.

• Proceedings of the Korea Concrete Institute Conference
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• 1994.10a
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• pp.285-290
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• Proceedings of the Korea Concrete Institute Conference
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• 1994.10a
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• pp.291-300
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• 1994

The influence of elevated temperatures on the mechanical properties of concrete is important for fire-resistance studies and also for understanding the behavior of containment vessel, such as nuclear reactor pressure vessels, during service and ultimate condition. The present study is to clarify the damage/deterioration of concrete structures that are subjected to high temperature exposure. To this end, comprehensive experiments are conducted. The major test variables are the peak temperatures, rate of temperature increase, and sustained duration at peak temperature. The results include weight loss residual compressive strength and stress-strain curve. From those results, residua compressive strength formula and stress-strain relationship are proposed.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.88-103
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• 1999

The purposes of this study are to understand the severe accident phenomena, to establish the simulation method for the experimental test, and to assess the current models in MELCOR for future improvement. This paper presents the results of the PHEBUS FPT1 post test analysis using MELCOR computer code, version 1.8.4. The entire PHEBUS facility has been modeled; the core, the primary circuit including the steam generator, and the containment vessel. Both the thermal hydraulic and the fission product behavior have been investigated. The code simulation results of the thermal hydraulic behavior show good agreement with the experimental data, The fission product release and transport are calculated using the CORSOR models in MELCOR code and the results will be compared with the experiment when the experimental data are available.

• Journal of the Korean Society of Safety
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• v.11 no.2
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• pp.33-41
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• 1996

Experiments were performed to study solidification of phase change material on a finned vertical tube when either conduction In the solid or natural convection in a liquid controls the heat transfer. The liquid was housed in a cylindrical containment vessel whose surface was maintained at a uniform, time-invariment temperature during a data run, and the solidification occurred at a finned and unfinned vertical tube positioned along the axis of the vassel. The phase change material(PCM) employed in this experiment is 99 percent pure n-Octacosan paraffin($C -{28}H_{58}/$). For conduction-controlled and convection-controlled solidification, the enhancement of the solidified mass rate due to finning is great when the solidified layer is thin and decreases as the layer grows thicker. It is studied that the latent energy($E_{\lambda}$) is the largest contributor to the total extracted energy($E_{\lambda} ＋ E_{sl}＋E_{s2}$) and the total extracted energy rate at a finned vertical tube is greater than that at a unfinned vertical tube.