• Journal of Mechanical Science and Technology
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• v.17 no.9
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• Steel and Composite Structures
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• v.29 no.2
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• pp.187-200
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• 2018

This paper handles the repairing of deficient square Concrete-Filled Steel-Tube (CFST) beams subject to bending through an experimental and numerical program. Eight square-CFST beams were tested. A 5-mm artificial notch was induced at mid-span of seven beams, four of them were repaired by using CFRP sheets and two were repaired by using GFRP sheets. The beam deflection, strain and ultimate moments were recorded. It was found that providing different cut-off points for the different layers of FRP sheets prohibited failure at termination points due to stress concentrations. Using different lengths of FRP sheets around the notch retarded crack propagation and prevented FRP rupture at the crack position. Finite element analysis was then conducted and the proposed FE model was verified against the recorded experimental data. The influence of various parameters as FRP sheet length, tensile modulus and the number of layers were studied. The moment capacity of damaged square-CFST beams was improved up to 77.6% when repaired by using four layers of CFRP, however, this caused a dramatic decrease in beam deflection. U-wrapping of notched-CFST beam with 0.75 of its length provided a comparable behaviour as wrapping the full length of the beam.

• Korean Journal of Head & Neck Oncology
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• v.35 no.2
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• pp.27-30
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• 2019

Parathyroid adenoma can cause extracapsular bleeding. In 1934, Capps first reported a case of massive hemorrhage secondary to rupture of a parathyroid adenoma. Recently, we experienced a 73-year-old female presented with pharyngeal discomfort and extensive ecchymosis over the neck without history of trauma. Endoscopic investigation revealed submucosal hemorrhage in the posterior wall of the hypopharynx. CT scan and ultrasonography demonstrated the presence of a mass below the left thyroid lobe. Serum calcium level was normal and PTH level was elevated. We underwent left thyroidectomy and parathyroidectomy 2 weeks later from first visit. During the operation, hypopharyngeal mucosa was teared and it was treated with pharyngostoma formation and L-tube feeding. We report a rare case of normocalcemic parathyroid adenoma with spontaneous hemorrhage and propose the proper management period with a literature review.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1361-1374
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• 2009

Due to the need of data for a Human Reliability Analysis (HRA), a number of data collection efforts have been undertaken in several different organizations. As a part of this effort, a human error analysis that focused on a set of simulator records on a Steam Generator Tube Rupture (SGTR) scenario was performed by using the Human Event Repository and Analysis (HERA) system. This paper summarizes the process and results of the HERA analysis, including discussions about the usability of the HERA system for a human error analysis of simulator data. Five simulated records of an SGTR scenario were analyzed with the HERA analysis process in order to scrutinize the causes and mechanisms of the human related events. From this study, the authors confirmed that the HERA was a serviceable system that can analyze human performance qualitatively from simulator data. It was possible to identify the human related events in the simulator data that affected the system safety not only negatively but also positively. It was also possible to scrutinize the Performance Shaping Factors (PSFs) and the relevant contributory factors with regard to each identified human event.

• Nuclear Engineering and Technology
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• v.44 no.4
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• pp.393-404
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• 2012

After the Fukushima nuclear accident in 2011, there has been increasing concern regarding severe accidents in nuclear facilities. Severe accident scenarios are difficult for operators to monitor and identify. Therefore, accurate prediction of a severe accident is important in order to manage it appropriately in the unfavorable conditions. In this study, artificial intelligence (AI) techniques, such as support vector classification (SVC), probabilistic neural network (PNN), group method of data handling (GMDH), and fuzzy neural network (FNN), were used to monitor the major transient scenarios of a severe accident caused by three different initiating events, the hot-leg loss of coolant accident (LOCA), the cold-leg LOCA, and the steam generator tube rupture in pressurized water reactors (PWRs). The SVC and PNN models were used for the event classification. The GMDH and FNN models were employed to accurately predict the important timing representing severe accident scenarios. In addition, in order to verify the proposed algorithm, data from a number of numerical simulations were required in order to train the AI techniques due to the shortage of real LOCA data. The data was acquired by performing simulations using the MAAP4 code. The prediction accuracy of the three types of initiating events was sufficiently high to predict severe accident scenarios. Therefore, the AI techniques can be applied successfully in the identification and monitoring of severe accident scenarios in real PWRs.

• Journal of ILASS-Korea
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• v.9 no.4
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• pp.24-30
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• 2004

Feedwater flowing tube side of number 5 high pressure feedwatrr heaters was heated by extracting steam from high pressure turbine and draining water from moisture separators and number 6 high pressure feedwater heaters and supplied into steam generators. Because the extracting steam from the high pressure turbine is two phase fluid of high temperature, high pressure, and high speed and flows to inverse direction after impinging to impingement baffle. the shell wall of the number 5 high pressure feedwater heater may be affected by flow accelerated corrosion. On May 14, 1999, Point Beach Nuclear Plant (PBNP) with operating at full power experienced a steam leak from rupture of shell side of number 4B feedwater heater. Also, d domestic nuclear power plant experienced a severe wall thinning of shell side of number 5A and 5B feedwater heaters. This paper describes the fluid mixing analysis study using PHOENICS code in order to get at the root of the shell wall thinning of the feedwater heaters. The sections included in the fluid mixing analysis model are around the number 5h feedwater heater shell including the extracting pipeline. To identify the relation between the local velocities and wall thinning. the local velocities according to the analysis results were compared with the distribution of the shell wall thickness by ultrasonic test.

• Journal of Chest Surgery
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• v.22 no.5
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• pp.788-793
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• 1989

We have observed 501 cases of spontaneous pneumothorax from January 1981 to June 1989 at the Department of Thoracic and Cardiovascular Surgery, Chonbuk National University Hospital. Of these, 57 patients have undergone thoracotomy to treat the pneumothorax after closed thoracostomy. These 57 patients were based on this retrospective clinical analysis, and the results were as follows: The ratio of male to female was 4.2:1 in male predominance and the old aged patients, over 50 years old, occupied 47.3% of all patients. Primary spontaneous pneumothorax was 19 cases and secondary spontaneous pneumothorax was 38 cases. The underlying pathology in secondary spontaneous pneumothorax was tuberculosis emphysema and chronic obstructive pulmonary disease in 35 cases. The indications of thoracotomy were persistent air leakage in 23 cases recurrent pneumothorax in 21 cases, inadequate expansion in 13 cases. Rupture of bullae or blebs were most frequent operative and pathologic findings in persistent air leakage group and recurrent pneumothorax group. In inadequate expansion group, predominant finding was destructive lung lesion. Bullectomy and/or bullae ligation was most effective procedures in 36 cases [63%] for operative management of spontaneous pneumothorax. Duration of preoperative and postoperative chest tube indwelling day was 13.35 days and 8.05 days in persistent pneumothorax group, 8.92 days and 7.77 days in recurrent pneumothorax group, 13.23 days and 10.21 days in inadequate expansion group.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.723-730
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• 2019

Acquiring instrumentation signals generated from nuclear power plants (NPPs) is essential to maintain nuclear reactor integrity or to mitigate an abnormal state under normal operating conditions or severe accident circumstances. However, various safety-critical instrumentation signals from NPPs cannot be accurately measured on account of instrument degradation or failure under severe accident circumstances. Reactor vessel (RV) water level, which is an accident monitoring variable directly related to reactor cooling and prevention of core exposure, was predicted by applying a few signals to deep neural networks (DNNs) during severe accidents in NPPs. Signal data were obtained by simulating the postulated loss-of-coolant accidents at hot- and cold-legs, and steam generator tube rupture using modular accident analysis program code as actual NPP accidents rarely happen. To optimize the DNN model for RV water level prediction, a genetic algorithm was used to select the numbers of hidden layers and nodes. The proposed DNN model had a small root mean square error for RV water level prediction, and performed better than the cascaded fuzzy neural network model of the previous study. Consequently, the DNN model is considered to perform well enough to provide supporting information on the RV water level to operators.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.402-409
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• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.51-58
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• 2021

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.