• Nuclear Engineering and Technology
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• 제48권2호
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• pp.448-456
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• 2016

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies themass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The $H_2O$ content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of $CO_2$ necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3491-3496
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• 2007

If a molten core is released from a reactor vessel into a reactor cavity during a severe accident, an important safety issue of coolability of the molten core from top-flooding and concrete ablation due to a molten core concrete interaction (MCCI) is still unresolved. The released molten core debris would attack the concrete wall and basemat of the reactor cavity, which will lead to inevitable concrete decompositions and possible radiological releases. In a OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests were performed to secure the data for cooling the molten core spread out at the reactor cavity and for the 2-D long-term core concrete interaction (CCI). The tests included not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with a prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. The paper deals with the transient simulations on the CCI-2 test by using a severe accident analysis code, CORQUENCH, which was developed at Argonne National Laboratory (ANL). Similar simulations had been already per for me d by using MELCOR 1.8.5 code. Unlike the MELCOR 1.8.5, the CORQUENCH includes a melt eruption mode I and a newly developed water ingression model based on the water ingression tests under the OECD/MCCI project. In order to adjust the geometrical differences between the CCI-2 test (rectangular geometry) and the simulations (cylindrical geometry), the same scaling methodology as used in the MELCOR simulation was applied. For the direct comparison of the simulation results, the same inputs for the MELCOR simulation were used. The simulation results were compared with the previous results by using MELCOR 1.8.5.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.268-271
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• 2008

If a reactor vessel is failed to retain a molten corium in a postulated severe accident, the molten corium is released outside the reactor vessel into a reactor cavity. The molten corium would attack the concrete wall and basemat of the reactor cavity, which may lead to inevitable concrete decompositions and possible radiological releases. In the OECD/MCCI project, a series of tests were performed to secure the data for cooling the molten corium spread out at the reactor cavity and for the long-term CCI (Core Concrete Interaction). Also, a MCCI (Molten Core Concrete Interaction) analysis code, CORQUENCH was upgraded at Argonne National Laboratory with embedding the new models developed for the tests. This paper deals with analyses of MCCI at plant scale under the conditions of top flooding using the upgraded CORQUENCH code. The modeling approach is briefly summarized first, followed by presentation of a validation calculation that illustrates the predicative capability of the modeling tool. With this background in place, the model is then used to carry out a parametric set of scoping calculations that define approximate coolability envelopes for the LCS (Limestone Common Sand) concrete that has been evaluated in the OECD/MCCI project.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.922-932
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• 2024

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl₃ fuel and NaCl-KCl-MgCl₂ coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

• BMB Reports
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• 제37권6호
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• pp.676-683
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• 2004

The conformational properties of hydrophobic core variant ubiquitin (Val26 to Ala mutation) in an acidic solution were studied. The intrinsic tryptophan fluorescence emission spectrum, far-UV and near-UV circular dichroic spectra, the fluorescence emission spectrum of 8-anilinonaphthalene-1-sulfonic acid in the presence of V26A ubiquitin, and urea-induced unfolding measurements indicate this variant ubiquitin to be in the partially folded molten globule conformation in solution at pH 2. The folding kinetics from molten globule to the native state was nearly identical to those from the unfolded state to the native state. This observation suggests that the equilibrium molten globule state of hydrophobic core variant ubiquitin is an on-pathway folding intermediate.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.495-501
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• 1997

A preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only neutronic aspects of the accident were considered, independent of the accident scenario, and efforts were made to estimate the quantity of molten fuel which must be ejected out of the core to assure a sub-critical state after the accident. Two types of parameters were examined : characteristic parameters of molten core such as geometry, molten pool type (homogenized or stratified), fuel temperature, environment, and relative parameters to normal core such as core size(small or large), and fuel type (oxide, nitride, metal). The first type of parameters was found to intervene more directly in the recriticality risk than the second type of parameters.

• 한국주조공학회지
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• 제17권4호
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• pp.402-410
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• 1997

Developing a salt core for squeeze casting process, two different salt cores(pure salt core and mixed salt core) were fabricated and investigated. Pure salt core was composed of 100% NaCl and mixed salt core was made by mixtures of NaCl with MgO(1%), $Na_2B_4O_7$(2%), and talc(1%) as a binder or a strengthening agent. Salt cores were compacted to various theoretical density, heat treated, and then squeeze-cast with molten Al alloy(AC8A). The compression strength of salt cores were measured and the squeeze-cast products were examined for shape retention, infiltration of molten metal into the cores, and microstructures. The shape of salt core compacted at above 75% of the theoretical density was maintained stably. The higher theoretical density of salt cores gave higher compression strength, and the compression strength of mixed salt core was higher than that of pure salt core. Namely at 90% theoretical density, the compression strength of mixed salt core was $6.3 kg/mm^2$, compared to $4.6 kgmm^2$ for pure salt core. At a squeeze casting pressure of $1000 kg/cm^2$, molten Al alloy was infiltrated into pure salt core of under 85% of the theoretical density. At squeeze casting pressure of $1000 kg/cm^2$, only mixed salt core above 90% of the theoretical density were valid, but the shape of the core was altered in the case of pure salt core at 90% of theoretical density. A key factor for developing a salt core for squeeze casting process was estimated as the ultimate compressive strength of salt core.

• Nuclear Engineering and Technology
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• 제32권5호
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• pp.425-432
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• 2000

It has been an essential part for the safety assessment of nuclear power plants to understand various phenomena associated with the molten core-concrete interaction(MCCI) under severe accidents. In this study, the severe accident analysis code MELCOR was used to simulate the MCCI experiments such as SWISS and SURC test series which had been performed in Sandia National Laboratories(SNL). The calculation results were compared with corresponding experimental data such as melt temperature, concrete ablation distance, gas generation rate, and aerosol release rate. Good agreements were observed between MELCOR calculation and experimental data. The melt pool was sustained within the range of high temperature and the concrete ablation occurred continuously. The gas generation and aerosol release were under the influence of melt temperature and overlying water pool, respectively.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4266-4273
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• 2023

In a severe reactor accident, a crust will form on the surface of the molten material during the core melting process. The crust will have a contact melting with the internal components of the reactor. In this paper, the contact melting process of the molten material on the austenitic stainless steel plate bundles is studied. The contact melting model of parabolic molten material on the plate bundles is proposed, and the rule and main effect factors of the contact melting are analyzed. The results show that the melting velocity is proportional to the slope of the paraboloid, the heat flux and the distance between two plates D. The influence of melt gravity and the plate width on melting velocity is negligible. The thickness of the molten liquid film is proportional to the heat flux and plate width, and it is inversely proportional to the gravity. With the increase of D, the liquid film thickness decreases at first and then increases gradually. The liquid film thickness has a minimum against D. When the width of the plate is small, the width of the plate is the main factor affecting the thickness of the liquid film. The parameters are coupled with each other. In a severe reactor accident, the wider internal components of reactor, which can increase the thickness of the melting liquid film and reduce the net input heat flux from the molten material to the components, are the effective measures to delay the melting process.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.437-443
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• 2024

The development and testing of inspection equipment is necessary for the safe deployment of advanced nuclear reactors. One proposed advanced reactor design is Westinghouse's lead-cooled fast reactor (LFR). In this paper, the process of achieving adequate wetting for an ultrasonic under-lead viewing system is discussed and results presented. Such a device would be used for inspection in the molten lead core during reactor outages. Wider tests into the wetting of various materials in molten lead at microscale were performed using electron microscopy. The possible mechanisms and kinetics for materials wetting in lead, particularly stainless steel and nickel, are proposed and discussed.