• Journal of the Korean Society of Safety
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• v.35 no.4
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• 한국전산유체공학회:학술대회논문집
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• 2009.04a
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• pp.290-297
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• 2009

A two-phase (gas and liquid) flow analysis solver, named CUPID, has been developed for a realistic simulation of transient two-phase flows in light water nuclear reactor components. In the CUPID solver, a two-fluid three-field model is adopted and the governing equations are solved on unstructured grids for flow analyses in complicated geometries. For the numerical solution scheme, the semi-implicit method of the RELAP5 code, which has been proved to be very stable and accurate for most practical applications of nuclear thermal hydraulics, was used with some modifications for an application to unstructured non-staggered grids. This paper is concerned with the effects of interpolation schemes on the simulation of two-phase flows. In order to stabilize a numerical solution and assure a high numerical accuracy, the second-order upwind scheme is implemented into the CUPID code in the present paper. Some numerical tests have been performed with the implemented scheme and the comparison results between the second-order and first-order upwind schemes are introduced in the present paper. The comparison results among the two interpolation schemes and either the exact solutions or the mesh convergence studies showed the reduced numerical diffusion with the second order scheme.

• Journal of Chest Surgery
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• v.10 no.2
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• pp.299-314
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• 1977

Clinical experience on 16 cases of open heart surgery under the extracorporeal circulation with mild or moderate hypothermia and partial hemodilution technique at the National Medical Center during the period from June 1976 to October 1977. Nine of sixteen were congenital heart disease and seven were acquired heart disease. The age of the patient ranged between 6 and 48 years. The body weight varied from 18.5kg to 60kg and body surface area 0. 79-1.70m2. The average priming volume of pump oxygenator was 2080 ml, which was consisted fresh ACD blood, buffered Hartmann`s solution, Mannitol, 50% dextrose in water and Vit. C. The average hemodilution rate was 27%. The average flow 2.3 L/min/m2 or 80 ml/min and the duration of perfusion varied from 31 min to 270 min with average of 107 min. The perfusion was carried out under the mild or moderate hypothermia using core cooling alone in 10 cases, core cooling and local myocardial cooling with $0-4^{\circ}C$ physiologic saline in 2 cases. From a hemodynamic point of view, the blood pressure dropped down around 80 mmHg after the initiation of perfusion follwed by increase to safety level and stable during the perfusion. The central venous pressure remained within normal limits. In most cases, hemoglobin and hematocrit decreased during and after the perfusion. Hemogiobin level was decreased, average of 20.6 %, hematocrit 18.6%, pletelets 55% postoperatively. Plasma hemoglobin increased moderately, from preperfusion average valve of 7.79 mg % to post-perfusion value of 54.7 mg %. Electrolytes changes during cardiopulmonary bypass showed definite hypokalemia but changes of Na, Ca were not definite. Arterial blood gas analysis during cardiopulmonary bypass suggested that the metabolic acidosis which was accompanied by respiratory alkalosis which was corrected postoperatively. As the opera tive complication, transient hemoglobinuria in 4 cases and neurological signs in 2 cases were all cured. There were 2 death cases and operative mortality rate was 12.5%.

• Journal of the Korean Society of Safety
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• v.37 no.5
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• pp.80-88
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• 2022

The Korean nuclear industry had developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code, which adopts a two-fluid, three-field model that is comprised of gas, continuous liquid and droplet fields and has the capability to simulate three-dimensional models. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for the accident management plan of a nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification is required for the separate and integral effect experiments. Therefore, the goal of this work is to verify the calculation capability of the SPACE code for multiple failure accidents. For this purpose, an experiment was conducted to simulate a Control Element Drive Mechanism (CEDM) break with a safety injection failure using the ATLAS test facility, which is operated by Korea Atomic Energy Research Institute (KAERI). This experiment focused on the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The results of the overall system transient response using the SPACE code showed similar trends with the experimental results for parameters such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it can be concluded that the SPACE code has sufficient capability to simulate a CEDM break with a safety injection failure accident.