• Nuclear Engineering and Technology
• /
• v.43 no.3
• /
• pp.279-286
• /
• 2011

The aging degradation and lifetime assessment of a domestic class 1E Ethylene-Propylene-Diene-Monomer (EPDM), which is a popular insulating elastomer for electrical cables in the nuclear power plants, were studied for equipment qualification verification under the Loss of Coolant Accident (LOCA) conditions. The specimens were acceleratively aged, underwent a LOCA environment, as well as tested mechanically, thermo-gravimetrically, and spectroscopically according to the American Society of the Testing of Materials (ASTM) procedures. The tensile test results revealed that the elongation at break gradually decreased with an increasing aging temperature. The lifetime of EPDM aged isothermally at $140^{\circ}C$ was 1,316 hours and reduced to 1,120 hours after experiencing the severe accident test. The activation energies of the elongation reduction were $1.10{\pm}0.196$ eV and $0.93{\pm}0.191$ eV before and after the LOCA condition, respectively. The TGA test results also showed that the activation energy of the aging decomposition decreased from 1.35 eV to 1.02 eV after undergoing the LOCA environment. Although the mechanical property changes were discernibly observed during the aging process, along with the LOCA simulation, the FT-IR analysis showed that the spectroscopic peaks and their intensities did not alter significantly. Therefore, it can be concluded that the degradation of the domestic class 1E EPDM due to aging can be tolerable, even in severe accident conditions such as LOCA, and thus it qualifies as a suitable insulating material for electrical cables in the nuclear power plants.

• Nuclear Engineering and Technology
• /
• v.45 no.7
• /
• pp.871-894
• /
• 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 Radiation Protection and Research
• /
• v.20 no.1
• /
• pp.53-64
• /
• 1995

The radiation fields following the large loss of coolant accident (LOCA) have been assessed for the vital areas in the service building of Wolsong 2, 3, and 4 nuclear power plants. The ORIGEN2 code was used in calculating the fission product inventories in the fuel. The source terms were based upon the activity released following the dual failure accident scenario, i.e., a LOCA followed by impaired emergency core cooling (ECC). Configurations of the reactor building, the service building, and the ECC system were constructed for the QAD-CG calculations. The dose rates and the time-integrated doses were calculated for the time period of upto 90 days after the accident. The results showed that the radiation fields in the vital access areas were found to be sufficiently low. Some areas however showed relatively high radiation fields that may require limited access.

• Nuclear Engineering and Technology
• /
• v.41 no.5
• /
• pp.691-708
• /
• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.

• Journal of Energy Engineering
• /
• v.25 no.4
• /
• pp.85-92
• /
• 2016

Since the power density of the VHTR(Very High Temperature Reactor) is lower, there is less possibility of core melt. VHTR has no risk of explosion caused by hydrogen generation when the loss of coolant accident occurs, which is another advantage. Along with safety benefit, it can be used as a process heat supplier near demand facilities because coolant temperature is very high enough to be used for industrial purpose. In this paper, we designed the primary system using VHTR and the secondary system providing electricity and process heat. Based on that 350 MW thermal reactor proposed by NGNP(Next Generation Nuclear Part), we developed conceptual model that the IHX(Intermediate Heat Exchanger) loop transports 300 MW thermal energy to the secondary system. In addition, we analyzed thermodynamic behavior and performed the efficiency analysis and optimization study depending on major parameters.

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2053-2068
• /
• 2023

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.

• 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.

• Journal of Energy Engineering
• /
• v.9 no.3
• /
• pp.261-268
• /
• 2000

A heat transfer correlation to predict the vaporization of a water droplet in highly superheated steam during a loss-of-coolant accident(LOCA) of a nuclear power plant is provided. Vaporization of liquid fuel or water droplets in superheated air or steam and subsequent interface heat transfer between a liquid droplet and superheated gas is typically correlated by way of a Nusselt number as a function of Reynolds number, Prantl number, and in some cases including mass transfer number. Presently available correlations and experimental data of the evaporation of liquid droplets in air or steam are analyzed and a new Nusselt number correlation is proposed taking Schmidt number into consideration in order to account for binary diffusion of the vapor as well, Nu$\_$f/(1+B)$\^$0.7/=2+0.53Sc$\_$f/$\^$-1/5/Re$\_$M/$\^$$\sfrac{1}{2}$/Pr$\_$f/$\^$$\sfrac{1}{3}$/ for which properties are evaluated at film condition except the density of Reynolds number evaluated at ambient condition. Diverse correlations for various combinations of liquid and gas species are put into single equation. The blowing correction factor of (1+B)$\^$0.7/ is confirmed appropriate, and a criterion to distinguish so-called high- and low-temperature condition of ambient gas is set forth.

• Nuclear Engineering and Technology
• /
• v.55 no.3
• /
• pp.850-860
• /
• 2023

An industrial control process multiple-input multiple-output (MIMO) coupled system is analyzed in this study as an example of a Loss of Coolant Accident (LOCA) simulation system. Ordinary control algorithms can complete the steady state of the control system and even reduce the response time to some extent, but the entire system still consumes a large amount of energy after reaching the steady state. So a multivariable decoupled energy-saving control method is proposed, and a novel energy-saving function (economic function, Eco-Function) is specially designed based on the active disturbance rejection control algorithm. Simulations and LOCA simulation system tests show that the Eco-function algorithm can cope with the uncertainty of the multivariable system's internal parameters and external disturbances, and it can save up to 67% of energy consumption in maintaining the parameter steady state.

• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.1045-1052
• /
• 2009

A calculation model is developed to predict the transient free surface flow on the containment floor following a loss-of-coolant accident (LOCA) of pressurized water reactors (PWR) for the use of debris transport evaluation. The model solves the two-dimensional Shallow Water Equation (SWE) using a finite volume method (FVM) with unstructured triangular meshes. The numerical scheme is based on a fully explicit predictor-corrector method to achieve a fast-running capability and numerical accuracy. The Harten-Lax-van Leer (HLL) scheme is used to reserve a shock-capturing capability in determining the convective flux term at the cell interface where the dry-to-wet changing proceeds. An experiment simulating a sudden break of a water reservoir with L-shape open channel is calculated for validation of the present model. It is shown that the present model agrees well with the experiment data, thus it can be justified for the free surface flow with accuracy. From the calculation of flow field over the simplified containment floor of APR1400, the important phenomena of free surface flow including propagations and interactions of waves generated by local water level distribution and reflection with a solid wall are found and the transient flow rates entering the Holdup Volume Tank (HVT) are obtained within a practical computational resource.