• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.352-359
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• 2020

The effects of hydrogen precipitation on the mechanical properties of Zircaloy-4 and Zirlo alloys were examined with uniaxial tensile tests at room temperature and at 400 ℃ and accompanying microstructural changes in the Zircaloy-4 and Zirlo alloy specimens were discussed. The elastic moduli of Zircaloy-4 and Zirlo alloys decreased with increasing hydrogen concentrations. Yield strengths of both materials tended to decrease gradually. The reductions of yield stress seems to be caused by the dissipation of yield point phenomena shown in stress-strain curves. Ultimate tensile strengths (UTS) of Zircaloy-4 and Zirlo slightly increased at low hydrogen contents, and then decreased when the concentrations exceeded 500 and 700 wppm, respectively. Uniform elongations were stable until 600 wppm and drops to 0% around 1400 wppm at room temperature.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.681-688
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• 2014

Hydride reorientation behaviors of PWR cladding tubes under typical interim dry storage conditions were investigated with the use of as-received 250 and 485ppm hydrogen-charged Zr-Nb alloy cladding tubes. In order to evaluate the effect of typical cool-down processes on the radial hydride precipitation, two terminal heat-up temperatures of 300 and $400^{\circ}C$, as well as two terminal cool-down temperatures of 200 and $300^{\circ}C$, were considered. In addition, two cooling rates of 2.5 and $8.0^{\circ}C/min$ during the cool-down processes were taken into account along with zero stress or a tensile hoop stress of 150MPa. It was found that the 250ppm hydrogen-charged specimen experiencing the higher terminal heat-up temperature and the lower terminal cool-down temperature generated the highest number of radial hydrides during the cool-down process under 150MPa hoop tensile stress, which may be explained by terminal solid hydrogen solubilities for precipitation, and dissolution and remaining circumferential hydrides at the terminal heat-up temperatures. In addition, the slower cool-down rate generates the larger number of radial hydrides due to a cooling rate-dependent, longer residence time at a relatively high temperature that can accelerate the radial hydride nucleation and growth.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.238-248
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• 2019

In the high-temperature and high-pressure irradiation environments, the multi-field coupling processes of hydrogen diffusion, hydride precipitation and mechanical deformation in Zircaloy cladding tubes occur. To simulate this hydrogen-induced complex behavior, a multi-field coupling method is developed, with the irradiation hardening effects and hydride-precipitation-induced expansion and hardening effects involved in the mechanical constitutive relation. The out-pile tests for a cracked cladding tube after irradiation are simulated, and the numerical results of the multi-fields at different temperatures are obtained and analyzed. The results indicate that: (1) the hydrostatic stress gradient is the fundamental factor to activate the hydrogen-induced multi-field coupling behavior excluding the temperature gradient; (2) in the local crack-tip region, hydrides will precipitate faster at the considered higher temperatures, which can be fundamentally attributed to the sensitivity of TSSP and hydrogen diffusion coefficient to temperature. The mechanism is partly explained for the enlarged velocity values of delayed hydride cracking (DHC) at high temperatures before crack arrest. This work lays a foundation for the future research on DHC.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.12 s.255
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• pp.1526-1533
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• 2006

The aim of this research was to confirm the existence of the thermal creep degradation by hydrides of Zr-2.5Nb pressure tube materials. Small punch creep tests were performed to obtain the relationship between a creep displacement and a loading period at $300^{\circ}C$. A creep stress and a creep strain rate were also converted from the previous results. The creep material constants and the creep stress exponents at the different hydride contents were compared. Finally the hydrides of the axial and circumferential section were observed using OM, SEM and TEM. The following conclusions were made: 1) The degradation of the thermal creep by hydrides was existed and it strongly depended on the hydride contents. 2) As the hydride contents were increased, the creep stress exponents (m) were also increased. 3) Even though the hydride was not precipitated in 50 ppm materials at $300^{\circ}C$, the degradation of thermal creep was found. Therefore, it was believed that this phenomenon strongly related to the hydride precipitation at room temperature.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.746-755
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• 2015

To investigate the effect of hydrogen and oxygen contents on hydride reorientations during cool-down processes, zirconium-niobium cladding tube specimens were hydrogen-charged before some specimens were oxidized, resulting in 250 ppm and 500 ppm hydrogen-charged specimens containing no oxide and an oxide thickness of $0.38{\mu}m$ at each surface. The nonoxidized and oxidized hydrogen-charged specimens were heated up to $400^{\circ}C$ and then cooled down to room temperature at cooling rates of $0.3^{\circ}C/min$ and $8.0^{\circ}C/min$ under a tensile hoop stress of 150 MPa. The lower hydrogen contents and the slower cooling rate generated a larger fraction of radial hydrides, a longer radial hydride length, and a lower ultimate tensile strength and plastic elongation. In addition, the oxidized specimens generated a smaller fraction of radial hydrides and a lower ultimate tensile strength and plastic elongation than the nonoxidized specimens. This may be due to: a solubility difference between room temperature and $400^{\circ}C$; an oxygen-induced increase in hydrogen solubility and radial hydride nucleation energy; high temperature residence time during the cool-down; or undissolved circumferential hydrides at $400^{\circ}C$.

• Journal of Ocean Engineering and Technology
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• v.23 no.2
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• pp.69-73
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• 2009

The objective of this study was to obtain a better understanding of the delayed hydride cracking (DHC) of Zr-2.5Nb alloy. The DHC model has some defects: first, it cannot explain why the DHC velocity (DHCV) becomes constant regardless of an applied stress intensity factor, even though the stress gradient is affected by the applied stress intensity factor at the notch tip. Second, it cannot explain why the DHCV has a strong dependence on the method of approaching the test temperature by a cool-down or a heating-up, even under the same stress gradient, and third, it cannot predict any hydride size effect on the DHC velocity. The DHC tests were conducted on Zr-2.5Nb compact tension specimens with the test temperatures reached by a heating-up method and a cool-down method. Crack velocities were measured in hydrided specimens, which were cooled from solution-treatment temperatures at different rates by being furnace-cooled, water-quenched, and liquid nitrogen-quenched. The resulting hydride size, morphology, and distributions were examined by optical metallography. It was found that fast cooling rates, which produce very finely dispersed hydrides, result in higher crack growth rates. This different DHC behavior of the Zr-2.5Nb tube with the cooling rate after a homogenization treatment is due to the precipitation of the $\gamma$-hydrides only in the water-quenched Zr-2.5Nb tube. This experiment will provide supporting evidence that the terminal solid solubility of a dissolution (TSSD) of $\gamma$-hydrides is higher than that of $\delta$-hydrides.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• 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.

• Analytical Science and Technology
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• v.18 no.5
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• pp.396-402
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• 2005

We try to look for optimum conditions of pre-reductants like L-Cysteine, KI and $FeSO_4$ when analyzing inorganic arsenic by using hydride generation-atomic absorption spectrometry, and run a comparative study of effect in the analysis of them. Also, we separated and analyzed only inorganic arsenic by using $H_2SO_4$-trap to eliminate organic arsenic which are MMA(monomethylarsonate) and DMA(dimethylarsinate). Under the conditions of mixture acid of 1.8 M HCl and 0.08 M $HNO_3$, arsenic standard solution of 20 ppb have more higher absorbance than without adding acid. In case of L-Cysteine, As(V) completely reduces into As(III) when 0.5 g of L-Cysteine is reacted more than 30 mins. in weak acid condition of approximately 0.07 M $HNO_3$ or HCl. In the event of KI, As(V) completely reduces into As(III) when 3 g of KI is reacted more than 1hour in acid condition of 0.8 M $HNO_3$. On the occasion of $FeSO_4$, the inside of tube is blocked by precipitation by mixture reaction of $NaBH_4$ and $Fe^{2+}$, therefore, comparing to other pre-reductants, reproducibility of efficiency of reducing As(V) to As(III) is low. To evaluate the accuracy of the analytical results, we use NIST SRM 1643C Trace Elements in Water ($82.1{\pm}1.2ng/mL$). The results are satisfactory.