• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1192-1197
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• 2003

The aim of this study is to investigate the hydride embrittlement when the LBB evaluation is carried out for the integrity of PHWR Pressure Tubes. The transverse tensile and CCT tests were performed at three hydrogen concentrations while the test temperatures were changed (RT to $300^{\circ}C$). The specimens were directly machined from the pressure tube retaining original curvature. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over $250^{\circ}C$. Using the DHC test results, the CCL and LBB time were calculated and compared. The hydride embrittlement behavior at the LBB evaluation was definitely showed.

• Corrosion Science and Technology
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• v.20 no.3
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• pp.118-128
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• 2021

The oxide layer in samples taken from an irradiated Zr-2.5Nb pressure tube from a CANDU-PHWR reactor was analyzed using electron probe microanalysis (EPMA). The examined tube had been exposed to temperatures ranging from 264 to 306 ℃ and a neutron fluence of 8.9 × 10²¹ n/cm² (E > 1 MeV) for the maximum 10 effective full-power years in a nuclear power plant. Measuring oxide layer thickness generally employs optical microscopy. However, in this study, analysis of the oxide layer from the irradiated pressure tube components was undertaken through X-ray image mapping obtained using EPMA. The oxide layer characteristics were analyzed by X-ray image mapping with 256 × 256 pixels using EPMA. In addition, the slope of the oxide layer was measured for each location. A particular advantage of this study was that backscattered electrons and X-ray image mapping were obtained at a magnification of 9,000 when 20 kV volts and 30 uA of current were applied to radiation-shielded EPMA. The results of this study should usefully contribute to the study of the oxide layer properties of various types of metallic materials irradiated by high radiation in nuclear power plants.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.5
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• pp.610-616
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• 2004

The aim of this study was to investigate the hydride embrittlement when the LBB evaluation was carried out for the integrity of PHWR Pressure Tubes. The transverse tensile and CCT toughness tests were performed at three hydrogen concentrations while the test temperatures were changed (RT to 30$0^{\circ}C$). Both the transverse tensile and the fracture toughness tests showed the hydrogen embitterment clearly at RT but this phenomenon was disappeared while the test temperature arrived at 25$0^{\circ}C$. Using the DHC test results, the CCL and LBB time were calculated and compared. The hydride embrittlement at the LBB evaluation made the LBB time short definedly. If the operating temperature, DHCV and LBB deterministic parameters such as A and m were known, LBB time could be estimated without the calculation of CCL.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.265.2-265
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• 2001

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.1
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• pp.60-67
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• 2016

A pressurized heavy water reactor (PHWR) designed to refuel in service produces the energy required by nuclear fission. The fuel channel consists of components such as a pressure tube which directly contacts the fuel and is a passage for the reactor coolant, a calandria tube which contacts the moderator and is rolled joint with calandria, and a spacer which is not to contact the pressure tube and a calandria tube. As the fuel channel is one of the most important equipments, it requires accurate and periodic inspections to assess the integrity of a reactor in accordance with CSA N285.4. A fuel channel inspection system is developed to inspect fuel channels during in-service inspection in Wolsong unit. In this paper, the results and considerations of a field test are presented in order to show the effectiveness of the developed fuel channel inspection system.

• International Journal of Control, Automation, and Systems
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• v.4 no.1
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• pp.87-95
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• 2006

The outlet feeder pipe thinning in a PHWR (Pressurized Heavy Water Reactor) is caused by a high pressure steam flow inside the pipe, which is a well known degradation mechanism called a FAC (Flow Assisted Corrosion). In order to monitor the degradation, the thickness of the outlet bends close to the exit of the pressure tube should be measured and analyzed at every official overhaul. This paper describes a mobile feeder pipe inspection robot that can minimize the irradiation dose to human workers by automating the measurement process. The robot can move by itself on the feeder pipe by using an inch worm mechanism, which is constructed by two gripper bodies that can fix the robot body on to the pipe, one extendable and contractible actuator, and a rotation actuator connected to the two gripper bodies to move forward and backward, and to rotate in a circumferential direction.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.369-377
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• 2003

Since Zr-2.5Nb pressure tubes have a high risk for the formation of blisters during their operation in pressurized heavy water reactors, there has been a strong incentive to develop a method for the non-destructive detection of blisters grown on the tube surfaces. However, because there is little mismatch in acoustic impedance between the hydride blisters and zirconium matrix, it is not easy to distinguish the boundary between the blister and zirconium matrix with conventional ultrasonic methods. This study has focused on the development of a special ultrasonic method, so called ultrasonic velocity ratio method for a reliable detection of blisters formed on Zr-2.5Nb pressure tubes. Hydride blisters were grown on the outer surface of the Zr-2.5Nb pressure tube using a cold finger attached to a steady state thermal diffusion equipment. To maximize a difference in the ultrasonic velocity in hydride blisters and the zirconium matrix, the ultrasonic velocity ratio of longitudinal wave to shear wave, $V_L/V_S$, has been determined based on the flight time of the longitudinal echo and reflected shear echo from the outer surface of the tubes. The feasibility of the ultrasonic velocity ratio method is confirmed by comparing the contour plots reproduced by this method with those of the blisters grown on the Zr-2.5Nb pressure tubes.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.78-83
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• 2003

The aim of this study is to investigate the effect of hydrogen embrittlement od Zr-2.5Nb CANDU pressure tube. The test were performed at three hydrogen contents for transverse tensile and CCT specimens while the test temperatures were changed (RT to 300$^{\circ}C$). The specimens were directly machined from the tube retaining original curvature using electric discharge machine. Both the transverse tensile and the fracture toughness tests showed the hydrogen embrittlement clearly at RT but this phenomenon was disappeared while the test temperature arrived over 250$^{\circ}C$. From the fracture toughness test, it was found that fracture toughness dJ/da was increased up to 200$^{\circ}C$ and then decreased.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.265-276
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• 2013

Pressure tubes made of Zr-2.5 wt% Nb alloy are important components consisting reactor coolant pressure boundary of a pressurized heavy water reactor, in which unanticipated through-wall cracks and rupture may occur due to a delayed hydride cracking (DHC). The Canadian Standards Association has provided deterministic and probabilistic structural integrity evaluation procedures to protect pressure tubes against DHC. However, intuitive understanding and subsequent assessment of flaw behaviors are still insufficient due to complex degradation mechanisms and diverse influential parameters of DHC compared with those of stress corrosion cracking and fatigue crack growth phenomena. In the present study, a deterministic flaw assessment program was developed and applied for systematic integrity assessment of the pressure tubes. Based on the examination results dealing with effects of flaw shapes, pressure tube dimensional changes, hydrogen concentrations of pressure tubes and plant operation scenarios, a simple and rough method for effective cooldown operation was proposed to minimize DHC risks. The developed deterministic assessment program for pressure tubes can be used to derive further technical bases for probabilistic damage frequency assessment.