• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3196-3206
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• 2021

For the Korean design of the PCCS (passive containment cooling system) in an innovative PWR, the overall thermal resistance around a condenser tube is dominated by the heat transfer coefficient of steam condensation on the exterior surface. It has been reported, however, that the calculated heat transfer coefficients by thermal-hydraulic system codes were much lower than measured data in separate effect tests. In this study, a new empirical model of steam condensation in the presence of a noncondensable gas was implemented into the MARS-KS 1.4 code to replace the conventional Colburn-Hougen model. The selected correlation had been developed from condensation test data obtained at the JERICHO (JNU Experimental Rig for Investigation of Condensation Heat transfer On tube) facility, and considered the effect of the Grashof number for naturally circulating gas mixture and the curvature of the condenser tube. The modified MARS-KS code was applied to simulate the transient response of the containment equipped with the PCCS to the large-break loss-of-coolant accident. The heat removal performances of the PCCS and corresponding evolution of the containment pressure were compared to those calculated via the original model. Various thermal-hydraulic parameters associated with the natural circulation operation through the heat transport circuit were also investigated.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1525-1531
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• 2019

The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4504-4517
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• 2023

After a loss of coolant accident (LOCA) in the prestressed concrete containment vessels (PCCVs) of nuclear power plants, the coupling of temperature and pressure can significantly affect the mechanical properties of the PCCVs. However, there is no consensus on how this coupling affects the failure mechanism of PCCVs. In this paper, a simplified finite element modeling method is proposed to study the effect of temperature and pressure coupling on PCCVs. The experiment results of a 1:4 scale PCCV model tested at Sandia National Laboratory (SNL) are compared with the results obtained from the proposed modeling approach. Seven working conditions are set up by varying the internal and external temperatures to investigate the failure mechanism of the PCCV model under the coupling effect of temperature and pressure. The results of this paper demonstrate that the finite element model established by the simplified finite element method proposed in this paper is highly consistent with the experimental results. Furthermore, the stress-displacement curve of the PCCV during loading can be divided into four stages, each of which corresponds to the damage to the concrete, steel liner, steel rebar, and prestressing tendon. Finally, the failure mechanism of the PCCV is significantly affected by temperature.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.11
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• pp.3260-3268
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• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• Journal of the Korean Society of Safety
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• v.30 no.3
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• pp.107-113
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• 2015

The goal of this paper is to investigate structural integrity factors of RMI(reflective metal insulation) to confirm the design requirements in nuclear power plant. Currently, a glass wool insulation is using now, but it will gradually be replaced with the reflective metal insulation maded by stainless steel plates. The main function of an insulation is to minimize a heat loss of vessel and pipes in RCS(reactor coolant system). It has to maintain structural a integrity in nuclear power plant life duration. In this study, the structural integrity analysis was carried out both multi-plate and outer shell plate by using a static analysis and experimental test. First, inner multi-plate has a self support structure for being air space. Because the effect of total static weight in multi-layer plate is low, a plate collapse possibility is not high. Considering optimum thin plate pressing process, it has to pre-check the basic physical properties. Second, the outer segment thickness and stiffener shape are verified by the numerical static analysis, and sample test for both type of panel and cylindrical pipe model.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.62-69
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• 2013

A probabilistic assessment code, PRO-LOCA ver. 3.7 which was developed in an international co-operative research program, PARTRIDGE was evaluated by conducting sensitivity analysis. The effect of some variables such as simulation methods (adaptive sampling, iteration numbers, weld residual stress model), crack features(Poisson's arrival rate, maximum numbers of cracks, initial flaw size, fabrication flaws), operating and loading conditions(temperature, primary bending stress, earthquake strength and frequency), and inspection model(inspection intervals, detectable leak rate) on the failure probabilities of a surge line nozzle was investigated. The results of sensitivity analysis shows the remaining problems of the PRO-LOCA code such as the instability of adaptive sampling and unexpected trend of failure probabilities at an early stage.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.315-326
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• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.373-380
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• 2014

Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.