• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.3
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• pp.731-740
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• 1995

The objective of this paper is to develop defect assessment technology for integrity evaluation of CANDU pressure tubes. In fracture mechanics analysis, three-dimensional and two-dimensional (line-spring model) finite element analyses were performed to obtain the stress intensity factor for axial and circumferential surface cracks. In leak before break (LBB) analysis, heat transfer analyses for through-wall cracks were performed by considering the cooling effect and the LBB application time was computed. It was shown that the analytical results obtained in this study provide less-conservative but accurate solution for defect assessment of CANDU pressure tubes.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.3
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• pp.311-317
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• 2004

This paper proposes engineering estimation equations for the maximum deflection of a cylinder subject to bending under elastic-plastic and elastic-creep conditions. Being based on the reference stress approach, the proposed equations are simple to use and can accommodate general tensile and creep behaviours. Validation against detailed 3-D FE results using actual stress-strain data and realistic creep-deformation data shows excellent agreement, which provides confidence in the use of the proposed equation. Based on the proposed equations, together with information on in-service inspection data, discussion is given how to estimate future time-dependent and time-independent deflection of the CANDU pressure tube. Thus the present result would be valuable information for integrity assessment of the CANDU pressure tube.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.479-490
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• 2005

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.

• Nuclear Engineering and Technology
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• v.30 no.3
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• pp.262-272
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• 1998

The experiences and causes of pressure tube cracking accidents in the CANDU reactors and the development of the fuel channel at AECL(Atomic Energy Canada Limited) have been described. Most of the accidents were caused by Delayed Hydride Cracking(DHC). In the cases of the Pickering units 3&4 and the Bruce unit 2, excessive residual stresses induced by an improper rolled joint process played a role in DHC. In the Pickering unit 2, cracks formed by contact between the pressure and calandria tubes due to the movement of the garter spring were the direct cause of the failure. To extend the life of a fuel channel, several R&D programs examining each component of the fuel channel have been carried out in Canada. For a pressure tube, the main concern is focused on changing the fabrication processes, e.g., increasing cold working rate, conducting intermediate annealing and adding a third element like Fe, V, and Cr to the tube material. In addition to them, chromium plating on the end fitting and increasing wall thickness at both ends of the calandria tube are considered. There has also been much interest in the improvement of fuel channel performance in our country and several development programs are currently under way.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.155-160
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• 2001

This paper describes a probabilistic fracture mechanics (PFM) analysis based on Monte Carlo (MC) simulation. In the analysis of CANDU pressure tube, it is necessary to perform the PFM analyses based on statistical consideration of flaw generation time. A depth and an aspect ratio of initial semi-elliptical surface crack, a fracture toughness value, delayed hydride cracking (DHC) velocity, and flaw generation time are assumed to be probabilistic variables. In all the analyses, degradation of fracture toughness due to neutron irradiation is considered. Also, the failure criteria considered are plastic collapse, unstable fracture and crack penetration. For the crack growth by DHC, the failure probability was evaluated in due consideration of flaw generation time.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.93-98
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• 2000

Unpredictable failures can occur due to the DHC (delayed hydride cracking) or the degradation of fracture toughness by hydride embrittlement in CANDU pressure tube which can result from the absorption of hydrogen or deuterium in the high temperature coolant. To investigate the hydride embrittlement of CANDU Zr-2.5Nb pressure tube, the transverse tensile test and the fracture toughness test were performed from room temperature to $300^{\circ}C$ using three different specimens which have an AR (As Received), 100, and 200 ppm hydrogen. As the amount of absorbed hydrogen was increased, the transverse yield strength and the ultimate tensile strength were also increased. In addition, as the test temperature became higher they were decreased linearly. While, at room temperature, the hydrogenbsorbed specimens represented the embrittlement which resulted in sudden decreasing of fracture toughness, the fracture characteristics became ductile such as AR specimen at high temperatures. Through the observation of fracture surface using SEM, it was found that the stress state of mixed mode could be related to the fissure which was believed to decrease the global fracture toughness.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.775-783
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• 1995

This study is to assess the effects of increasing wall thickness on the safety margin of pressure tube in operating and of lowering initial hydrogen concentration on the DHC growth in respect to the improvement of the reliability of pressure tube in CANDU reactors. The pressure tube with thicker wall of 5.2 mm shows much higher safety margin for flaw tolerance by 25％ than the current 4.2mmm tube. The thicker pressure tubes have a great benefit in LBB assessment including the initial crack depth at which DHC occurs, the crack length at onset of leaking and the available time for action. The resistance for the pressure tube ballooning at LOCA accident is also increased with the thicker tube. The calculations for Heq concentration after 20 years of operation as a function of wall thickness and initial hydrogen concentration show that the 5.2 mm nil thickness tube with 5 ppm initial hydrogen concentration is the most resistant to DHC. with the lower initial hydrogen concentration, TSS temperature for the precipitation or hydride decreases and the crack growth during cooldown reduces.

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.834-839
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• 2001

Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.422-422
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.247-247
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• 2002