• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.284-292
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• 2015

A numerical calculation with the commercial computational fluid dynamics code CFX-14.0 was conducted for a test facility simulating the Canadian deuterium uranium moderator thermal-hydraulic. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the Canadian deuterium uranium moderator circulating vessel, which is called a calandria tank, housing a matrix of horizontal rod bundles simulating calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the calandria system. In the present study, the full geometric details of the calandria tank are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.72-77
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• 2010

During the period of reinforcement work for the licensing renewal of CANDU NPP, the fuel channels, Calandria tubes and feeders of CANDU Reactor are replaced. The remote visual inspection of Calandria internal is also performed during the period of reinforcement work. This period is a unique opportunity to inspect the inside of the Calandria. The visual inspection for the Calandria vessel and its internals of Wolsong NPP Unit 1 was performed by Nuclear Engineering & Technology Institute(NETEC) of KHNP. To perform this inspection, NETEC developed equipment applied new technology such as the synchronization of 3D CAD, automatic alignment and control system. The inspection confirmed that the Calandria integrity of Wolsong NPP Unit 1 is perfect.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.57-61
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• 2008

During the period of retubing work for the licensing renewal, the fuel channels, calandria tubes and feeders of CANDU Reactors will be replaced, and calandria visual examination will be performed. This period is a unique opportunity to inspect the inside of the calandria. The visual inspection for the calandria vessel and its internals of Wolsong NPP is scheduled for confirming the calandria integrity. The first visual inspection for the calandria is planned in Pt. Lepreau led by AECL. The visual inspection for Wolsong NPP, led by NETEC(Nuclear Engineering & Technology Institute) of KHNP, will employ 3D laser scanner and 3D CAD Mock-up for the first time in the world, in addition to a conventional video camera. The inspection system is composed of a robot with the 3D laser scanner, a video camera and a hardness meter.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.321-322
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• 2004

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.90-100
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• 2021

Molten corium-concrete interaction (MCCI) is one of the most important phenomena that can lead to the potential hazard of late containment failure due to basemat penetration during a severe accident. In this study, MCCI analytical models of the CORQUENCH code were prepared through verification calculations of several experiments, which had been performed using concrete types similar to those of the calandria vault floor in CANDU-6 plants. The behaviors of thermal-hydraulic variables related to MCCI phenomena were analyzed under the conditions of dry floor and water flooding during the severe accident stemming from a hypothetic station blackout. Uncertainty analyses on the ablation depth were also carried out. It was estimated that the concrete ablation was not interrupted due to the continuous MCCI process under the dry condition but was terminated within 24 hours under the water flooding condition. It was confirmed that the water flooding as a mitigating action was effective to achieve the quenching and thermal stabilization of the melt discharged from the calandria vessel, showing that the present models are capable of reasonably simulating MCCI phenomena in CANDU-6 plants. This study is expected to provide the technical bases to the accident management strategy during the late-phase severe accidents.

• Journal of Energy Engineering
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• v.21 no.4
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• pp.419-426
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• 2012

A transient, three-dimensional, two-phase flow analysis code, CUPID, has been developed in KAERI. In this work, we performed a preliminary analysis using the CUPID code to investigate the thermal-hydraulic behavior of the moderator in the Calandria vessel of a CANDU reactor. At first, we validated the CUPID code using the three experiments that were performed at Stern Laboratories Inc. To avoid the complexity to generate computational mesh around the Calandria tube bundles, a porous media approach was applied for the region. The pressure drop in the porous media zone was modeled by an empirical correlation. The results of the calculations showed that the CUPID code can predict the mixed flow pattern of forced and natural convection inside the Calandria vessel very well. Thereafter, the analysis was extended to a two-phase flow condition. Also, the local maximum temperature in the Calandria vessel was plotted as a function of the injection flow rate, which may be utilized to predict the local subcooling margin.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.95-105
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• 2014

The CUPID code has been developed for a transient, three-dimensional, two-phase flow analysis at a component scale. It has been validated against a wide range of two-phase flow experiments. Especially, to assess its applicability to single- and two-phase flow analyses in the Calandria vessel of a CANDU nuclear reactor, it was validated using the experimental data of the 1/4-scaled facility of a Calandria vessel at the STERN laboratory. In this study, a preliminary thermal-hydraulic analysis of the CANDU reactor moderator tank using the CUPID code is carried out, which is based on the results of the previous studies. The complicated internal structure of the Calandria vessel and the inlet nozzle was modeled in a simplified manner by using a porous media approach. One of the most important factors in the analysis was found to be the modeling of the tank inlet nozzle. A calculation with a simple inlet nozzle modeling resulted in thermal stratification by buoyance, leading to a boiling from the top of the Calandria tank. This is not realistic at all and may occur due to the lack of inlet flow momentum. To improve this, a new nozzle modeling was used, which can preserve both mass flow and momentum flow at the inlet nozzle. This resulted in a realistic temperature distribution in the tank. In conclusion, it was shown that the CUPID code is applicable to thermal-hydraulic analysis of the CANDU reactor moderator tank using the cost-effective porous media approach and that the inlet nozzle modeling is very important for the flow analysis in the tank.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.275-276
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• 2004