• Nuclear Engineering and Technology
• /
• v.55 no.4
• /
• pp.1518-1527
• /
• 2023

A method was proposed for in-situ evaluating the thickness and resistivity of the oxide/hydroxide film formed on the surface of aluminum alloy exposed to sump water formed in the containment after a loss-of-coolant accident. The evaluation entailed fitting a model for the film impedance, which has film thickness and other variables describing the resistivity profile of the film along its thickness direction as fitting parameters, to the practically measured electrochemical impedance data. The obtained resistivity profiles implied that the films formed at pHs_25℃ 7, 8, 9, 10, and 11 all had a duplex structure; compared to the outer layer in contact with the solution, the inner layer of the film had a much higher resistivity and was inferred to be denser and provide most of the protectiveness of the film. Both the thickness and the total resistance of the film decreased with the increasing solution pH_25℃, suggesting that the films formed in more alkaline solutions had less protectiveness against corrosion, consistent with the increasing aluminum alloy corrosion rates previously identified.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.16 no.1
• /
• pp.37-41
• /
• 2020

The Nuclear containment building is a main safety-related structure that performs shielding and conservation functions to prevent highly radioactive materials from leakage to the outside environment in the case of various environmental conditions and postulated accidents. The containment building contains a reactor, steam generator, pressurizer, tank, reactor coolant system, auxiliary system and engineering safety system, and is designed so that highly radioactive materials above the limits specified in 10 CFR 100 do not escape to the outside environment in the case of LOCA(Loss of Coolant Accident) for instance. The containment metal liner plate(CLP) is a carbon steel plate with a nominal plate thickness of 6 mm, which functions as a mold for the wall and dome of the containment building when concrete is filled, fulfills airtightness to prevent leakage of seriously radioactive materials. In recent years, backside corrosion was found on the liner plate in some domestic nuclear power plants. The main cause of backside corrosion was unfilled concrete. In this paper, an inspection technique of assessing filling suitability for CLP backside concrete is developed. Results show that the validity of inspection technique for CLP backside concrete using vibration sensor is successfully verified.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.4003-4013
• /
• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

• Journal of the Korean Society of Safety
• /
• v.31 no.6
• /
• pp.129-134
• /
• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Nuclear Engineering and Technology
• /
• v.24 no.1
• /
• pp.86-97
• /
• 1992

The purpose of this paper is to evaluate the direct vessel injection design from a pressurized thermal shock(PTS) viewpoint for the Combustion Engineering System 80+ A break of the main steam line from zero power and a 0.05 ft$^2$small break loss-of-coolant accident (LOCA) from full power were selected as the potential PTS events. In order to investigate the stratification effects in the reactor downcomer region, the fluid mixing analysis was performed using the COMMIX-IB code for steam line break and using the REMIX code for 0.05 ft$^2$small break LOCA. The stress distributions within the reactor vessel walls experiencing the pressure and the temperature transients were calculated using the OCA-P code for both events. The results of the analysis showed that a small break LOCA without decay heat presented the greatest challenge to the vessel, however, there is no crack initiation through end-of-life of the vessel with consideration of decay heat.

• Nuclear Engineering and Technology
• /
• v.53 no.4
• /
• pp.1236-1249
• /
• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

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

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.29 no.12
• /
• pp.654-662
• /
• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• Nuclear Engineering and Technology
• /
• v.36 no.1
• /
• pp.24-35
• /
• 2004

In order to analyze the hydrogen distribution during a severe accident in the APR1400 containment, GASFLOW II was used. For the APR1400 NPP, a hydrogen mitigation system is considered from the design stage, but a fully time-dependent, three-dimensional analysis has not been performed yet. In this study GASFLOW code II is used for the three-dimensional analysis. The first step to analysis involving hydrogen behavior in a full containment with the GASLOW code is to generate a realistic geometry model, which includes nodalization and modeling of the internal structures such as walls, ceilings and equipment. Geometry modeling of the APR1400 is conducted using GUI program by overlapping the containment cut drawings in a graphical file format on the mesh view. The total number of mesh cells generated is 49,476. And the calculated free volume of the APR1400 containment by GASFLOW is almost the same as the value from the GOTHIC modeling. A hypothetical SB-LOCA scenario beyond design base accident was selected to analyze the hydrogen behavior with the hydrogen mitigation system. The source of hydrogen and steam for the GASFLOW II analysis is obtained from a MAAP calculation. Combustion pressure and temperature load possibilities within the compartments used in the GOTHIC analysis are studied based on the Sigma-Lambda criteria. Finally the effectiveness of HMS installed in the APR1400 containment is evaluated from the point of severe accident management

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