• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3264-3274
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• 2021

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1345-1354
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• 2019

Clogging of sump strainers that filter the recirculation water in containment after a loss-of-coolant accident (LOCA) seriously impedes the continued cooling of nuclear reactor cores. In experiments examining the corrosion of aluminium alloy 6061, a common material in containment equipment, in borated solutions simulating the water chemistry of sump water after a LOCA, we found that Fe-bearing intermetallic particles, which were initially buried in the Al matrix, were progressively exposed as corrosion continued. Their cathodic nature $vis-{\grave{a}}-vis$ the Al matrix provoked continuous trenching around them until they were finally released into the test solution. Such particles released from Al alloy components in a reactor containment after a LOCA will be transported to the sump entrance with the recirculation flow and trapped by the debris bed that typically forms on the strainer surface, potentially aggravating strainer clogging. These Fe-bearing intermetallic particles, many of which had a rod or thin strip-like geometry, were identified to be mainly the cubic phase ${\alpha}_c-Al(Fe,Mn)Si$ with an average size of about $2.15{\mu}m$; 11.5 g of particles with a volume of about $3.2cm^3$ would be released with the dissolution of every 1 kg 6061 aluminium alloy.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.902-907
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• 2023

To understand the eutectic reaction mechanism and the relocation behavior of the core debris is indispensable for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (B₄C) and stainless-steel (SS). The influence of the existence of carbon on the B₄C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B₄C-SS samples, a new layer was formed between B₄C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe₂B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples. All samples observed the Cr-rich domain and Fe and Ni-rich domain after the reaction. These domains might be formed during the solidifying process.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Journal of Korea Foundry Society
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• v.20 no.6
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• pp.377-385
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• 2000

Al matrix composites as the most promising MMCs can be expected to be excellent engineering materials in the nearest future. So as to improve material properties of composite, many manufacturing processes have been developed. Among them, squeeze casting process which offers fine microstructure and near-net-shape is one of the most successful MMCs manufacturing processes. But, in case of with subsieve size particles (under 44 ${\mu}m$), it is very difficult to homogeneously distribute particles in matrix of Al matrix composite by various casting processes, including squeeze casting used so far. Duplex process which was developed in previous study was used to distribute the particle of subsieve size more homogeneously in matrix of Al matrix composite. Microstructures, wear and heat resistance characterization of Al-Si-Cu-Mg-(Ni)/SiCp manufactured by duplex process were examined to clarify the effect of manufacturing conditions, particle size of reinforcement and alloying elements. Al matrix composites reinforced with SiCp(10 ${\mu}m$) have the lowest wear amount among composites reinforced with 3 ${\mu}m$, 5 ${\mu}m$ and 10 ${\mu}m$ SiCp. The wear amount of Al matrix composites with 10 wt.% SiCp(3, 5, 10 ${\mu}m$) was decreased according to the increase of the sliding speed because abrasive wear takes place at high sliding speed of 4m/s and worn debris with block type occurs at low sliding speed of 1m/s. As for heat resistance, it is made clear that remarkable heat resistance property can be obtained by addition of Ni element in Al matrix composites.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1206-1218
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• 2016

This article proposes a highly reliable power and communication system that guarantees the protection of essential instruments in a nuclear power plant under a severe accident. Both power and communication lines are established with not only conventional wired channels, but also the proposed wireless channels for emergency reserve. An inductive power transfer system is selected due to its robust power transfer characteristics under high temperature, high pressure, and highly humid environments with a large amount of scattered debris after a severe accident. A thermal insulation box and a glass-fiber reinforced plastic box are proposed to protect the essential instruments, including vulnerable electronic circuits, from extremely high temperatures of up to $627^{\circ}C$ and pressure of up to 5 bar. The proposed wireless power and communication system is experimentally verified by an inductive power transfer system prototype having a dipole coil structure and prototype Zigbee modules over a 7-m distance, where both the thermal insulation box and the glass-fiber reinforced plastic box are fabricated and tested using a high-temperature chamber. Moreover, an experiment on the effects of a high radiation environment on various electronic devices is conducted based on the radiation test having a maximum accumulated dose of 27 Mrad.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3631-3640
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• 2022

Uranium mill tailing ponds (UMTPs) are risk source of debris flow and a critical source of environmental U and Rn pollution. The technology of microbial induced calcium carbonate precipitation (MICP) has been extensively studied on reinforcement of UMTs, while little attention has been paid to the effects of MICP on U & Rn release, especially when incorporation of metakaolin and bacillus subtilis (MBS). In this study, the reinforcement and U & Rn immobilization role of MBS -MICP solidification in different grouting cycle for uranium mill tailings (UMTs) was comprehensively investigated. The results showed that under the action of about 166.7 g/L metakaolin and ~50% bacillus subtilis, the solidification cycle of MICP was shortened by 50%, the solidified bodies became brittle, and the axial stress increased by up to 7.9%, and U immobilization rates and Rn exhalation rates decrease by 12.6% and 0.8%, respectively. Therefore, the incorporation of MBS can enhance the triaxial compressive strength and improve the immobilization capacity of U and Rn of the UMTs bodies solidified during MICP, due to the reduction of pore volume and surface area, the formation of more crystals general gypsum and gismondine, as well as the enhancing of coprecipitation and encapsulation capacity.

• Journal of Astronomy and Space Sciences
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• v.33 no.2
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• pp.127-135
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• 2016

Inactive space objects are usually rotating and tumbling as a result of internal or external forces. KOREASAT 1 has been inactive since 2005, and its drift trajectory has been monitored with the optical wide-field patrol network (OWL-Net). However, a quantitative analysis of KOREASAT 1 in regard to the attitude evolution has never been performed. Here, two optical tracking systems were used to acquire raw measurements to analyze the rotation period of two inactive satellites. During the optical campaign in 2013, KOREASAT 1 was observed by a 0.6 m class optical telescope operated by the Korea Astronomy and Space Science Institute (KASI). The rotation period of KOREASAT 1 was analyzed with the light curves from the photometry results. The rotation periods of the low Earth orbit (LEO) satellite ASTRO-H after break-up were detected by OWL-Net on April 7, 2016. We analyzed the magnitude variation of each satellite by differential photometry and made comparisons with the star catalog. The illumination effect caused by the phase angle between the Sun and the target satellite was corrected with the system tool kit (STK) and two line element (TLE) technique. Finally, we determined the rotation period of two inactive satellites on LEO and geostationary Earth orbit (GEO) with light curves from the photometry. The main rotation periods were determined to be 5.2 sec for ASTRO-H and 74 sec for KOREASAT 1.