• Nuclear Engineering and Technology
• /
• v.55 no.2
• /
• pp.623-639
• /
• 2023

The specific configuration of the safety systems in VVER-1000/V320 reactors allows a comprehensive study of the Loss of Coolant Accident (LOCA). In the present paper, a verification of the success criteria of the event trees headers for the medium and large break LOCA sequences is conducted. A detailed TRACEV5P5 thermal-hydraulic model of the reactor has been developed, including all safety systems. When analyzing the results of all sequences, some conservatism is observed in certain specific configurations as the success criterion of some headers is not consistent with the classic PSA level 1. Therefore, new proposals for the LOCA event trees are performed based on a reconfiguration of LOCA break ranges and the use of the expanded event trees approach.

• Microbiology and Biotechnology Letters
• /
• v.37 no.4
• /
• pp.377-382
• /
• 2009

Viral safety is an important prerequisite for clinical preparations of all biopharmaceuticals derived from plasma, cell lines, or tissues of human or animal origin. To ensure the safety, implementation of multiple viral clearance (inactivation and/or removal) steps has been highly recommended for manufacturing of biopharmaceuticals. Of the possible viral clearance strategies, Ultraviolet-C (UVC) irradiation has been known as an effective viral inactivating method. However it has been dismissed by biopharmaceutical industry as a result of the potential for protein damage and the difficulty in delivering uniform doses. Recently a continuous flow UVC reactor (UVivatec) was developed to provide highly efficient mixing and maximize virus exposure to the UV light. In order to investigate the effectiveness of UVivatec to inactivate viruses without causing significant protein damage, the feasibility of the UVC irradiation process was studied with a commercial therapeutic protein. Recovery yield in the optimized condition of $3,000\;J/m^2$ irradiation was more than 98%. The efficacy and robustness of the UVC reactor was evaluated with regard to the inactivation of human immunodeficiency virus (HIV), hepatitis A virus (HAV), bovine herpes virus (BHV), bovine viral diarrhea virus (BVDV), porcine parvovirus (PPV), bovine parvovirus (BPV), minute virus of mice (MVM), reovirus type 3 (REO), and bovine parainfluenza virus type 3 (BPIV). Non enveloped viruses (HAV, PPV, BPV, MVM, and REO) were completely inactivated to undetectable levels by $3,000\;J/m^2$ irradiation. Enveloped viruses such as HIV, BVDV, and BPIV were completely inactivated to undetectable levels. However BHV was incompletely inactivated with slight residual infectivity remaining even after $3,000\;J/m^2$ irradiation. The log reduction factors achieved by UVC irradiation were ${\geq}3.89$ for HIV, ${\geq}5.27$ for HAV, 5.29 for BHV, ${\geq}5.96$ for BVDV, ${\geq}4.37$ for PPV, ${\geq}3.55$ for BPV, ${\geq}3.51$ for MVM, ${\geq}4.20$ for REO, and ${\geq}4.15$ for BPIV. These results indicate that UVC irradiation using UVivatec was very effective and robust in inactivating all the viruses tested.

• Fire Science and Engineering
• /
• v.26 no.5
• /
• pp.35-40
• /
• 2012

Numerical investigation was performed to study on the soot formation characteristics in the WSR according to the CO addition. Ethylene and pure air were used as a fuel and an oxidizer, respectively, and three different equivalence ratios (2.0, 2.5, 3.0) were used in the calculation. The resulted CO mole fraction of 10 % CO addition showed the maximum value in spite of the least CO supply. This means that the conversion of CO to soot and other carbon compounds is weakened under incipient soot formation. The soot volume fraction was decreased with increasing the CO addition because the important species for soot formation such as pyrene and acetylene, were decreased with the addition of CO. When the equivalence ratio was 2.5, the soot volume fraction shows the highest value, which results from the contribution of fuel rich condition and reacting temperature. Furthermore, surface growth rate and species concentrations justified the HACA mechanism for soot formation.

• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.1053-1064
• /
• 2009

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.37 no.5
• /
• pp.649-655
• /
• 2013

During welding, the dissimilar metal butt welds of nuclear piping are typically subjected to repair welding in order to eliminate defects that are found during post-weld inspection. It has been found that the repair weld can significantly increase the tensile residual stress in the weldment, and therefore, accurate estimation of the weld residual stress due to repair weld, especially for dissimilar metal welds using Ni-based alloy 82/182 in nuclear components, is of great importance in order to assess susceptibility to primary water stress corrosion cracking. In the present study, the stress distributions of dissimilar metal butt welds in nuclear reactor piping subjected to repair weld were investigated based on detailed nonlinear finite element analyses. Particular emphasis was placed on the variation of the stress distribution in the dissimilar metal butt weld according to the finite element welding analysis sequence for the repair welding process.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.20 no.9
• /
• pp.2894-2900
• /
• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.

• Journal of KIISE:Software and Applications
• /
• v.32 no.12
• /
• pp.1178-1191
• /
• 2005

In this paper, we propose a synthesis method of software fault tree from software requirements specification written in NuSCR formal specification language. The software fault tree, proposed in this paper, reflects requirements on both structure and behavior and it is an integrated form. The software fault tree can be used for analyzing safety in the view of structure and behavior. We propose templates for each components in NuSCR specification language and a synthesis method of software fault tree using the templates. The research was applied into the main trip logic of the reactor protection system of ARP1400, the Korean next generation nuclear reactor system, developed by KNICS. And we evaluate feasibility of our approach through this case study.

• Journal of Energy Engineering
• /
• v.26 no.3
• /
• pp.71-77
• /
• 2017

In the event of loss of coolant accident (LOCA) and station black out (SBO) in the primary system of a nuclear reactor, the coolant water should be injected to reactor coolant system (RCS) without any intervention of operators or active components. To satisfy the requirements, hybrid safety injection tank (Hybrid SIT) was suggested by Korea Atomic Energy Research Institute (KAERI). The pressure equalizing time of Hybrid SIT is an important parameter to determine the timing of coolant injection. To predict the pressure equalizing time of the Hybrid SIT, a separate effect test facility was constructed and sensitivity tests were conducted in various conditions. The main parameter determining the pressure equalizing time was obtained from separate effect test (SET) results. The wall of condensation on the inner wall of SIT and direct contact condensation on the water surface affected to the pressure equalizing time very much. In this study, the effect of each condensation phenomena on pressure equalizing time was quantitatively analyzed from results of SET and a prediction method of pressure equalizing time was proposed.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.613-624
• /
• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• The KSFM Journal of Fluid Machinery
• /
• v.16 no.4
• /
• pp.35-43
• /
• 2013

Most of the nuclear power plants(NPPs) adopts pressure difference type flow meters such as venturi and orifice meters for the measurement of feedwater flow rates to calculate reactor thermal power. However, corrosion products in the feedwater deposits on the flow meter by fouling as operating time goes. These effects lead to severe errors in the flow indication and then determination of reactor thermal power. The averaging BDFT, which has developed by Yun et al., has a potentiality to minimize this problem thanks to its inherent measurement principle. Therefore, it is expected that the averaging BDFT can replace the venturi meter for the feedwater pipe of steam generator of NPPs. The present work compares the amplification factor, K, based on CFD calculation against the K obtained from experiments in order to confirm whether a commercial CFD code can be applicable to the evaluation of characteristic for the averaging BDFT. In addition to this, the simulations to take into account of fouling effect are also carried out by rough wall option. The results show that the averaging BDFT is a promising flow meter for the accurate measurement of flow rates in the fouling condition of the NPPs.