• Proceedings of the Korea Institute of Fire Science and Engineering Conference
• /
• 2008.04a
• /
• pp.222-225
• /
• 2008

To prove a lot of technical difficulties related to the safety management of a railroad fire effectively, we design for the full-scale fire test facility of the railroad vehicle. It will be consist of major 3 part - duct system with smoke cleaning system, measuring section and gas analysis system. The CFD simulation was also carried out to design of the hood and duct system optimization. The results will be help for basic research of the railroad fire safety.

• Nuclear Engineering and Technology
• /
• v.39 no.4
• /
• pp.287-298
• /
• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.793-803
• /
• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1537-1546
• /
• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.

• Journal of the Korea Safety Management & Science
• /
• v.13 no.1
• /
• pp.133-142
• /
• 2011

Overhead facility design problem(OFDP) is one of the shortest rectilinear flow network problem(SRFNP)[4]. Genetic algorithm(GA), artificial immune system(AIS), population management genetic algorithm (PM) and greedy randomized adaptive search procedures (GRASP) were introduced to solve OFDP. A path matrix formed individual was designed to represent rectilinear path between each facility. An exchange crossover operator and an exchange mutation operator were introduced for OFDP. Computer programs for each algorithm were constructed to evaluate the performance of algorithms. Computation experiments were performed on the quality of solution and calculations time by using randomly generated test problems. The average object value of PM was the best of among four algorithms. The quality of solutions of AIS for the big sized problem were better than those of GA and GRASP. The solution quality of GRASP was the worst among four algorithms. Experimental results showed that the calculations time of GRASP was faster than any other algorithm. GA and PM had shown similar performance on calculation time and the calculation time of AIS was the worst.

• Nuclear Engineering and Technology
• /
• v.28 no.5
• /
• pp.440-451
• /
• 1996

With emphasis on safety, this study addresses for better design condition for the cooling system in a wet-type interim spent fuel storage facility, using a probabilistic safety assessment method. To incorporate the design renovation into the design phase, a simple approach is proposed. By taking the cooling system of a reference design, a fault tree analysis was performed to identify the weak point of the considered system, and then basic factors for design renovation were defined. A total of 21 design alternatives were selected through the combination of the basic factors. Finally, the optimum design alternative for the cooling system is derived by means of the cost and effect analysis based on the estimated cost, system reliability and assumed probabilistic safety criteria. With the assumption that the failure frequency of at-reactor spent fuel cooling system compiles with probabilistic safety criteria for the interim spent fuel cooling system, it was shown that the optimum alternative should have l00% cooling loop redundancy with one pump per cooling loop and a cleanup system installed separately from the main loop. Furthermore, it also should be classified into safety system. The result of this study can be used as a useful basis to identify factors of safety concern and to establish design requirements in the future. The method also can be applied for other nuclear facilities.

• Journal of the Korea Institute of Building Construction
• /
• v.20 no.3
• /
• pp.253-260
• /
• 2020

With increasing interest in an underground-type ammunition storage facility, several design results have been provided recently. However, since not only experts in the tunnel but also military persons in charge of ammunition have not fully understood the safety distance standard, reliable design results are not being produced. In this study, the effective design method of an underground-type ammunition storage facility was provided by analyzing the current safety distance standard. First, the critical safety distances that dominate the size of construction site for underground-type ammunition storage facilities were evaluated, which are the layout of chambers and the configuration of the entrances. Then, the decreasing effect of inter-chamber distance was studied according to the rock type and the storage density of ammunition. In addition, the method of designing tunnels with parallel lines and two-floors was considered for arranging more chambers while complying with the safety distance standards. In particular, numerical simulations were carried out to determine the satisfaction of the safety distance standards when an underground-type ammunition storage facility is composed of two-floor and the decreasing effect of inter-chamber distance according to the inner explosive pressure reduction. Finally, the method to adjust the size of entrances and the path of pressure were studied for decreasing the safety distance at the entrance.

• Nuclear Engineering and Technology
• /
• v.49 no.2
• /
• pp.411-425
• /
• 2017

Using a probabilistic safety assessment, a risk evaluation framework for an aircraft crash into an interim spent fuel storage facility is presented. Damage evaluation of a detailed generic cask model in a simplified building structure under an aircraft impact is discussed through a numerical structural analysis and an analytical fragility assessment. Sequences of the impact scenario are shown in a developed event tree, with uncertainties considered in the impact analysis and failure probabilities calculated. To evaluate the influence of parameters relevant to design safety, risks are estimated for three specification levels of cask and storage facility structures. The proposed assessment procedure includes the determination of the loading parameters, reference impact scenario, structural response analyses of facility walls, cask containment, and fuel assemblies, and a radiological consequence analysis with dose-risk estimation. The risk results for the proposed scenario in this study are expected to be small relative to those of design basis accidents for best-estimated conservative values. The importance of this framework is seen in its flexibility to evaluate the capability of the facility to withstand an aircraft impact and in its ability to anticipate potential realistic risks; the framework also provides insight into epistemic uncertainty in the available data and into the sensitivity of the design parameters for future research.

• Journal of the Korea Safety Management & Science
• /
• v.25 no.2
• /
• pp.29-38
• /
• 2023

Social welfare facilities are used by a wide range of local residents, including vulnerable populations such as the elderly, children, and people with disabilities. During emergencies like fires, confusion can arise as these individuals try to evacuate. Evacuation simulation results have shown that utilizing evacuation systems based on specific evacuation scenarios can significantly decrease the time required for evacuation compared to general evacuation procedures. By anticipating potential fires based on changes in social and facility environments, appropriate evacuation scenarios can be developed and applied to evacuation systems, thus contributing to the safety and security of individuals during emergencies. In conclusion, for social welfare facilities that serve a large number of people, it is necessary to expand the focus on performance-based design depending on the size of the facility, and to continuously develop and train for appropriate evacuation scenarios that align with changing facility environments.

• Nuclear Engineering and Technology
• /
• v.55 no.2
• /
• pp.792-799
• /
• 2023

Non-reactor nuclear facilities are increasing remarkably in Korea combined with advanced technologies such as life and space engineering, and the diversification of the nuclear industry. However, the absence of a basic classification guideline related to the design of non-reactor nuclear facilities has created confusion whenever related projects are carried out. In this paper, related domestic and international technical guidelines are reviewed to present the classification criteria of non-reactor nuclear facilities in Korea. Based on these criteria, the classification of structures, systems and components (SSCs) for safety controls is presented. Using the presented classification criteria, classification of a hot cell facility, a representative non-reactor nuclear facility, was performed. As a result of the classification, the hot cell facility is classified as the hazard category 3, accordingly, the safety class was classified as non-nuclear safety, the seismic category as non-seismic (RW-IIb), and the quality class as manufacturers' standards (S).