• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.4 s.235
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• pp.600-605
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• 2005

In this study, the ring tensile test at high temperature was suggested to evaluate the hoop tensile properties of small tube such as the cladding in the nuclear reactor Using the Arsene's ring model, the ring tensile test was performed and the test data were calibrated. From the result of the ring test with strain gauge and the numerical analysis with 1/8 model, LCRR(load-displacement conversion relationship of ring specimen) was determined. We could obtain the hoop tensile properties by means of applying the LCRR to the calibrated data of the ring tensile test. A few difference was observed in view of the shape of fractured surface and the fracture mechanism between at the high temperature and at the room temperature.

• Proceedings of the Optical Society of Korea Conference
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• 2000.02a
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• pp.214-215
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• 2000

As silica based optical fibers and preforms are processed at a high temperature, residual stresses are bulit in the strucure when cooled down to the room temperature. The magnitude of the residual stress depends on the difference in the thermal expansion coefficients between core and cladding glass as well as on the temperature difference. Residual stress distribution determines the intrinsic strength and could affect the long term reliability of optical fibers. And furthermore, stress can introduces anisotropy into optical fibers by photoelastic effects. The analysis of thermal stress has been intensively studied for multimode fibers$^{(1)}$ and the authors and co-wokers recently reported the stress distribution in a depressed inner cladding structure$^{(2)}$ . The compositions of the glass in the previous studies, however, have been restricted to conventional glass formers, such as GeO2, B2O3, P2O5, Fluorine. (omitted)

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.474-483
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• 2021

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters-burst stress, burst strain, burst temperature-during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• Composites Research
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• v.29 no.1
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• pp.40-44
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• 2016

Nuclear fuel cladding used in a nuclear power plant must possess superior oxidation resistance in the coolant atmosphere of high temperature/high pressure. However, as was the case for the critical LOCA (loss-of-coolant accident) accident that took place in the Fukushima disaster, there is a risk of hydrogen explosion when the nuclear fuel cladding and steam reacts dramatically to cause a rapid high-temperature oxidation accompanied by generation of a huge amount of hydrogen. Hence, an active search is ongoing for an alternative material to be used for manufacturing of nuclear fuel cladding. Studies are currently aimed at improving the safety of this cladding. In particular, ceramic-based nuclear fuel cladding, such as SiC, is receiving much attention due to the excellent radiation resistance, high strength, chemical durability against oxidation and corrosion, and excellent thermal conduction of ceramics. In the present study, cladding with $SiC_f/SiC$ protective films was fabricated using a process that forms a matrix phase by polymer impregnation of polycarbosilane (PCS) after filament-winding the SiC fiber onto an existing Zry-4 cladding tube. It is analyzed the oxidation and microstructure of the metal cladding with $SiC_f/SiC$ composite protective films using a drop tube furnace for thermal shock test.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.305-310
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• 1996

This paper concerns the cladding residual stresses in a reactor vessel induced during cooling from the manufacturing temperature down to room temperature Finite element results show that very large stress gradients are present at the interface corner and such stress singularity might lead to local yielding or cladding-base metal debonding.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.246-252
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• 2018

A series of simulated reactivity-initiated accident (RIA) tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC) test method. The EDC method reproduces fuel pellet-clad mechanical interaction (PCMI) conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm). The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.256-261
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

• Proceedings of the KWS Conference
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• 2001.10a
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• pp.65-68
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test(40$0^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test(40$0^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone