• 한국안전학회지
• /
• 제30권2호
• /
• pp.48-55
• /
• 2015

The number of collapse accidents relevant to form support since 2003 is 30 on the basis of statistical data from Ministry of Employment and Labor,. Total number of casualty was 138 (47 for deaths and 91 for injuries). The accident severity rate was high because the 4.6 casualties per one accident were occurred averagely although the incident rate was relatively low. Especially, one of form support members, the pipe supports have not been equipped adequately so that the accidents could have happened. In this regard, this study performed buckling test related to the effect of horizontal connection and slope in the single pipe which is one of typical pipe supports. The buckling load, which was estimated from the single pipe with the horizontal connectors theoretically and experimentally, was increased as more than 2 times compared to the buckling load obtained from the pipe supports without the connector. The buckling load was reduced as more than 26%, 34% slope of the single pipe comparing with 5% and 10% slope, respectively. Thus, the purpose of this study is to provide the guideline for installation and the maintenance of the pipe supports legally and institutionally to prevent the collapse accidents of the pipe supports.

• 한국유체기계학회 논문집
• /
• 제17권6호
• /
• pp.109-114
• /
• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Nuclear Engineering and Technology
• /
• 제46권6호
• /
• pp.817-824
• /
• 2014

To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon [2], showed almost same trend of physical properties.

• 한국기계가공학회지
• /
• 제19권1호
• /
• pp.43-48
• /
• 2020

At present, 24 nuclear power plants are in operation nationwide as the main power source responsible for about 27% of Korea's electricity, and five nuclear power plants are currently under construction. Issues of nuclear safety and reliability have always existed, but after the Fukushima accident, ensuring reliability has become an even more important issue for safety. Compared to other kinds of accidents, the initial response after a nuclear accident is more important than any other accident. Prior to accidents, it is important to be able to predict and judge the accident in advance for the sake of prevention. In this research, non-destructive inspection methods for existing pipe welds include radiographic, ultrasonic, magnetic particle practice, and liquid penetration testing. For this experiment, carbon steel pipes like that of the material used in nuclear pipes were adopted, and specimen welded to the flange (Flange) were manufactured. After testing, the weld specimen were not damaged through the infrared thermography (IRT) experiment. This study attempted to improve the safety of carbon steel pipes through a comparative analysis of finite element analysis.

• Journal of Trauma and Injury
• /
• 제29권1호
• /
• pp.33-36
• /
• 2016

Traffic accidents involving agricultural machinery have been decreasing, but mortality is still high due to a lack of safety devices such as seat belts. Furthermore, secondary damage, such as abdominal impalement injury caused by loaded materials, is more likely to occur, and this type of injury leads to a poor prognosis. Impalement with pipes is often more fatal than other penetrating injuries because the diameter of the pipe is usually larger in size than other loaded materials. We report a case of a 72-year-old man with secondary abdominal impalement injury caused by a scaffolding pipe following a traffic accident.

• Nuclear Engineering and Technology
• /
• 제53권12호
• /
• pp.3879-3891
• /
• 2021

A heat pipe residual heat removal system is proposed to be incorporated into the reactor driven subcritical (RDS) facility, which has been proposed by MIT Nuclear Reactor Laboratory for testing and demonstrating the Fluoride-salt-cooled High-temperature Reactor (FHR). It aims to reduce the risk of the system operation after the shutdown of the facility. One of the main components of the system is an air-cooled heat pipe heat exchanger. The alkali-metal high-temperature heat pipe was designed to meet the operation temperature and residual heat removal requirement of the facility. The heat pipe model developed in the previous work was adopted to simulate the designed heat pipe and assess the heat transport capability. 3D numerical simulation of the subcritical facility active zone was performed by the commercial CFD software STAR CCM + to investigate the operation characteristics of this proposed system. The thermal resistance network of the heat pipe was built and incorporated into the CFD model. The nominal condition, partial loss of air flow accident and partial heat pipe failure accident were simulated and analyzed. The results show that the residual heat removal system can provide sufficient cooling of the subcritical facility with a remarkable safety margin. The heat pipe can work under the recommended operation temperature range and the heat flux is below all thermal limits. The facility peak temperature is also lower than the safety limits.

• 한국안전학회지
• /
• 제30권2호
• /
• pp.41-47
• /
• 2015

Recently, a lot more disasters in the temporary structures happen because the stabilities of the temporary structures are disregarded by the reduction of the unit cost, using defective materials, the existing materials and so on. Pipe supports, which are one of the temporary structures, are basically used for the most constructing works such as buildings, bridges, plants and so on. In the most sites, adequate support installations of the pipe supports have not been performed although the presence of the guideline legally and institutionally. In this study, therefore, the collapse accidents of the pipe supports were investigated on the basis of theoretical analysis as well as the buckling tests by simulating the site support condition. Both the theocratical analysis and test results show that the buckling load in the fixed ends is at least 4 times larger than one in the pinned ends. This results will be utilized for safety assurance as well as accident prevention the in the field application.

• 상하수도학회지
• /
• 제27권6호
• /
• pp.679-687
• /
• 2013

Although stable and safe drinking water supply to the customers is a basic function of multi-regional water supply systems in Korea, most systems have their vulnerabilities in emergency time due to the branch-type. Application of connections from the other water supply system can provide a solutions for these tentative problems. This paper describes reduction planning of water supply accidents that can minimize a service interruption to customers in multi-regional water supply system by connecting pipe lines between local water supply systems in Mokpo city areas. The result of this study shows that Juam dam multi-regional water supply systems can cover all of the water shortage in southern parts of Jeonnam multi-regional water supply systems by transmitting water through connected pipes between local networks. This can be effective to supply water interactively in various contingencies, when a pipe line accident occurs in southern area of Jeonnam multi-regional water supply systems. On the contrary, southern area of Jeonnam multi-regional water supply systems can cover 99.5 %($62,500m^3/day$) of the water shortage of Juam dam multi-regional water supply systems when service interruptions caused by various pipe accidents occur in the system.

• Nuclear Engineering and Technology
• /
• 제37권6호
• /
• pp.537-556
• /
• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• 한국정밀공학회:학술대회논문집
• /
• 한국정밀공학회 2007년도 춘계학술대회 논문집
• /
• pp.113-114
• /
• 2007