• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.206-213
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• 2003

This study focuses on an understanding of the environmental effect on delayed hydride cracking velocity (DHCV) of CANDU Zr-2.5Nb tubes. To simulate DHC susceptibility of the Zr-2.5Nb tubes in reactor operating conditions, DHC tests were successfully carried out in pressurized water at 180 and $250^{\circ}C$ using a self-designed autoclave for the first time. Using 17 mm compact tension specimens electorlytically charged to 34 and 60 ppm H, 3 to 7 DHCV data were determined in water at both temperatures and compared to those determined in air that were already confirmed to be valid through a round robin test on DHCV of Zr-2.5Nb tubes sponsored by a IAEA coordinated research program. The pressurized water environment has little effect on DHCV of Zr-2.5Nb tube in water at both temperatures even though DHCV is slightly lower in water than that in air. The lower DHCV of the Zr-2.5Nb tube during short-term tests is discussed in viewpoint of the cooling rate from the peak temperature to the test temperature.

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.226-233
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• 1995

The wall thickness of a pressure tube is increased in order to reduce the probability of failure in a pressure tube of CANDU type reactor. It is presented here that the variation of wall thickness changes stress, hydrogen concentration and delayed hydride cracking in Zr-2.5Nb pressure tube. When the wall thickness is increased from 4.2 mm to 5.2 mm, the stress exerted on the tube and the deuterium taken up during operation are reduced by 19％. Further, the calculated allowable depth of the surface flaw over which delayed hydride cracking(DHC) is susceptible increases by 50％. DHC initiation is controlled by the stress and by the hydrogen concentration in the pressure tube. The results are therefore very significant in such a respect that increased wall thickness may reduce DHC initiation. Ac the wall thickness increases the hydrostatic tension will increase. Its impact on the acceleration of the crack growth rate of DHC deserves further studies.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.1133-1138
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• 2003

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration for blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• Journal of the Korean Society for Nondestructive Testing
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• v.22 no.1
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• pp.59-64
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• 2002

A FEM model of the remote-field eddy current effect is presented for zirconium-2.5 percent niobium(Zr-2.5%Nb) nuclear reactor pressure tubes to demonstrate the important electromagnetic field phenomena. This model is applied to evaluate the optimal operating frequency and detector position. There are many ambiguous experimental results connected with this technique. Finite element calculations can be used in the interpretation of these experimental results even though the electromagnetic fields measured in the remote-field technique are very small.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.2
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• pp.189-195
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• 2004

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration fur blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• 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.

• 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.

• Proceedings of the KIEE Conference
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• 2001.07b
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• pp.690-692
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• 2001

A FEM model of the remote-field eddy current effect is presented for zirconium-2.5percent niobium(Zr-2.5%Nb) nuclear reactor pressure tubes to demonstrate the important electromagnetic field. Phenomena that describe this effect. This model is applied to evaluate the optimal operating frequency and detector position. There are many ambiguous experimental results connected with this technique. Finite element calculations can be used in the interpretation of these experimental results even though the electromagnetic fields measured in the remote-field technique are very small.

• Journal of the Korean Society for Nondestructive Testing
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• v.34 no.3
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• pp.254-259
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• 2014

A pressurized heavy water reactor (PHWR) core has 380 fuel channels contained and supported by a horizontal cylindrical vessel known as the calandria, whereas a pressurized water reactor (PWR) has only a single reactor vessel. The pressure tube, which is a pressure-retaining component, has a 103.4 mm inside diameter ${\times}$ 4.19 mm wall thickness, and is 6.36 m long, made of a zirconium alloy (Zr-2.5 wt% Nb). This provides support for the fuel while transporting the $D_2O$ heat-transfer fluid. The simple tubular geometry invites highly automated inspection, and good approach for all inspection. Similar to all nuclear heat-transfer pressure boundaries, the PHWR pressure tube requires a rigorous, periodic inspection to assess the reactor integrity in accordance with the Korea Nuclear Safety Committee law. Volumetric-based nondestructive evaluation (NDE) techniques utilizing ultrasonic and eddy current testing have been adopted for use in the periodic inspection of the fuel channel. The eddy current testing, as a supplemental NDE method to ultrasonic testing, is used to confirm the flaws primarily detected through ultrasonic testing, however, eddy current testing offers a significant advantage in that its ability to detect surface flaws is superior to that of ultrasonic testing. In this paper, effectiveness of flaw detection and the depth sizing capability by eddy current testing for the inside surface of a pressure tube, will be introduced. As a result of this examination, the ET technique is found to be useful only as a detection technique for defects because it can detect fine defects on the surface with high resolution. However, the ET technique is not recommended for use as a depth sizing method because it has a large degree of error for depth sizing.