• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.3
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• pp.510-519
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• 2001

The tensile and fracture toughness tests have been conducted to investigate the degradations of mechanical properties induced mainly by neutron irradiations in Zr-2.5Nb CANDU pressure tube materials operated in Wolsung Unit-1. the tests were performed at room, 150, 200, 250, 300 $\^{C}$ for the irradiated and unirradiated specimens in hot cell. The specimens were directly machined from the tube retaining original curvature using specially designed electric discharge machine(EDM). From the tensile tests of the irradiated specimens, it was found that tensile strength was increased and total elongation was decreased compared to those of the unirradiated ones. The active voltages in the fracture toughness tests for the irradiated showed the discontinuous abrupt increases caused by crack jumping in lower temperature. In the crack resistance curves we found the stable crack growth in the unirradiated, whereas the unstable and three crack growth stages in the irradiated specimens due to the accumulated irradiation defects. The various fracture characteristic values in the irradiated are remarkably lower than those of the unirradiated. Through the fractography, we found in the irradiated that smaller dimple and shorter fissures than the unirradiated, and that the fractured surface had three regions that were flat, transition and slant/shear area. These can explain the difference in the crack growth characteristic values of the irradiated and the unirradiated ones.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.188-195
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• 2002

To investigate the degradation of mechanical properties induced mainly by neutron irradiation, the tensile tests were conducted from room temperature to 300\\`c using the irradiated and the unirradiated Zr-2.5Nb pressure tube materials. The irradiated longitudinal and transverse specimens were collected from the coolant inlet, middle, and outlet parts of M-11 tube which had been operated in Wolsung CANDU Unit-1 and exposed to different operating temperatures and irradiation fluences. The different tensile behavior was characterized not by the fluences of irradiation but by the tensile loading direction. The transverse specimen showed the higher strength and lower elongation than those of the longitudinal one. It was believed that these phenomena resulted from the microstructure anisotropy caused by the extrusion process. The increased strength hardening and decreased elongation embrittlement of the irradiated material were compard to those of the unirradiated one. While the tensile strength of the inlet was higher than that of the outlet, the elongation of the inlet was lower than that of outlet. Considering the operation condition, it was proposed that the operating temperature could be a more effective parameter than the irradiation fluence for long-time life. Through the TEM observation, it was found that while the a-type dislocation density was increased, the c-type dislocation was not changed in the irradiated. The fact that the higher dislocation density was sequentially distributed over the inlet, the middle, and the outlet parts was consistent with the distribution of the tensile strength.

• Transactions of the Korean Society of Mechanical Engineers
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• v.13 no.2
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• pp.221-228
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• 1989

The objective of this paper is to develop a numerical technique for analyzing crack driving forces in multi-phase materials. The analysis was based on finite element method coupled with a virtual crack extension technique which is known as the most efficient tool in computational fracture mechanics analysis. The modified J-integral method, proposed by Miyamoto and Kikuchi for the analysis of dual-phase material was carried out by subtracting the J-values for contours surrounding each phase boundary from the J-values for overall contour. It was shown that the proposed numerical procedure, based on the modified J-integral coupled with a virtual crack extension technique, can be used as an effective numerical tool for determining crack driving forces in multi-phase materials.

• The Journal of Korean Institute for Practical Engineering Education
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• v.2 no.1
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• pp.126-134
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• 2010

Material Testing Method subject of Metal Materials, which is one of curriculum of engineering major high school, is not properly updated even though both science and technology are advancing rapidly. Especially, the need of fractographic technique, which can analyze the fracture surface with the advanced testing equipment and technique, is increasing. The importance and the analyzing method of Fractography are proved by the characteristic analysis of pressure tube fracture surface of HWPR. In the future, in order to increase the professional knowledge and their pride of engineering major high school student, these advanced subject should be included in the new crriculum of engineering major high school. This activity to involve the advanced subject should be carried out voluntary by the first line teacher before the responsible body.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.11
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• pp.2305-2311
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• 2002

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate fur the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.

• Journal of Radiation Protection and Research
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• v.35 no.2
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• pp.57-62
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• 2010

Maintenance on the water chamber of steam generator, the change of pressurizer heater, the removal of pressure tube feeder, and so on during outage in nuclear power plants (NPPs) has a likelihood of high radiation exposure to whole body of workers even short time period due to the high radiation exposure rates. In particular, it is expected that hands would receive the highest radiation exposure because of its contact with radiation materials. In this study, field tests on extremity dose assessment of radiation workers for contact works with high radiation exposure were conducted during the maintenance periods in Korean pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs). In this field test, radiation workers were required to wear additional TLDs on the back and wrist, and an extremity dosimeter on fingers including a main TLD on the chest, while performing maintenance. As a result, it was found that the equivalent dose for fingers was distributed in the fixed range of deep dose and the equivalent dose for wrists.

• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.358-373
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• 1995

The Heat Transport system loop stability of CANDU-6 reactor using the CANFLEX fuel bundle was studied. The Thermal-hydraulic behavior of CANFLEX fuel bundle is similar to the conventional 37-element fuel bundle since the reactor power and the frictional pressure drop through the fuel channel is almost the same each other, Mounter the CANFLEX fuel bundle gives higher critical channel power and more homogeneous enthalpy distributions in the subchannels than 37-element fuel bundle. The SOPHT modelling or the CANFLEX fuel bundle and the Reactor outlet Header(ROH) interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. Without the ROH interconnection line the Heat Transport system loop using 43-element fuel bundle is unstable like the current 37-element fuel bundle. With the ROH interconnection line, however, the Heat Transport system is stable within $\pm$1％ of nominal flow. In the Heat Transport system loop stability point of view for Wolsong-1 plant therefore, the CANFLEX fuel loading is considered to be acceptable.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.13 no.3
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• pp.1255-1262
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• 2012

Look into the nuclear power plant of Wolsong currently, it is controlled in order to required operating pressure with PI controller. PI controller has a simple structure and satisfy design requirements to gain setting. However, It is difficult to control without changing the gain from produce changes in parameters such as loss of the valves and the pipes. To solve these problems, the dynamic change of the PI controller gain, or to compensate for the PI controller output is desirable to configure the controller. The aim of this research and development in the parameter variations can be controlled to a stable controller design which is reduced an error and a vibration. Proposed PI/NN control techniques is the PI controller and the neural network controller that combines a parallel and the neural network controller part is compensated output of the controller for changes in the parameters were designed to be robust. To directly evaluate the controller performance can be difficult to test in real processes to reflect the characteristics of the process. Therefore, we develope the simulator model using the real process data and simulation results when compared with the simulated process characteristics that showed changes in the parameters. As a result the PI/NN controller error and was confirmed to reduce vibrations.