• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.1-9
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• 2000

Delayed hydride cracking (DHC) velocity and threshold stress intensity factor for DHC ($K_{IH}$) tests in the radial direction on M11 pressure tube material in Wolsong unit 1 were carried out following the Atomic Energy Canada Limited (AECL) standard test procedure in order to identify the effect of undercooling on DHCV and to acquire the $K_{IH}$ data. The results showed that $K_{IH}$ 's were 8.8$\pm$0.8 MPa√m in the back offcut and 11.4$\pm$0.7 MPa√m in the front offcut. The fact that $K_{IH}$ in the front offcut is about 20% higher than that in the back offcut is attributed to the microstructural difference between the materials of the front and back ends. $K_{IH}$ 's in M11 pressure tube appeared to be higher than the values from the tubes made of double melted ingot reported earlier. This can be interpreted by the fact that very small amounts of Chlorine (Cl) and Phosphorus (P) are contained in the ingot and that the content of the harmful elements in the M11 pressure tube is equivalent to that made of a quadruple melting process. DHC velocities at 25$0^{\circ}C$ in the front offcut in the radial direction are measured to be 5~8$\times$10$^{-8}$ m/s. The results show that the prior thermal history change the DHC velocity significantly. This effect was confirmed by the experiment of undercooling prior to the DHC tests.DHC tests.

• 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.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.306-312
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• 2018

There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to $400^{\circ}C$ for high-burnup (>45 GWd/mtU) fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at $400^{\circ}C$. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.824-831
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• 2024

In order to investigate the occurring processes during long-term dry storage of spent fuel assemblies, a joined project called SPIZWURZ, between the Karlsruhe Institute of Technology and the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), was started. Aim of the SPIZWURZ project is the determination and quantification of the influence of texture and elastic strain on diffusion and solubility of hydrogen in three different zirconium alloys used in western Europe during a long-term cooling transient (1 K/d) starting at 400 ℃. The strain in the cladding of an irradiated spent fuel rod shall be measured. Models predicting the formation of radial oriented hydrides will be validated, improved, and implemented in the GRS fuel rod performance code TESPA-ROD. This paper describes the SPIZWURZ project and already obtained first results.