• Title/Summary/Keyword: Cladding oxide layer

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DYNAMIC CHARGE CARRIER TRANSPORT BEHAVIORS IN ZIRCONIUM OXIDE FOR NUCLEAR CLADDING MATERIALS

  • IL-KYU PARK;SANG-SEOK LEE;YONG KYOON MOK;CHAN-WOO JEON;HYUN-GIL KIM
    • Archives of Metallurgy and Materials
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    • v.65 no.3
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    • pp.1063-1067
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    • 2020
  • Dynamic charge carrier transport behavior in the zirconium (Zr) oxide was investigated based on the frequency-dependent capacitance-voltage (C-V) and temperature-dependent current-voltage (I-V) measurements. The Zr oxide was formed on the ZIRLO and newly developed zirconium-based alloy (NDZ) by corrosion in the PWR-simulated loop at 360℃. The corrosion test for 90 days showed that the NDZ exhibits better corrosion resistance than ZIRLO alloy. Based on the C-V measurement, dielectric constant values for the Zr oxide was estimated to be 11.28 and 11.52 for the ZIRLO and NDZ. The capacitance difference between low and high frequency was larger in the ZIRLO than in the NDZ, which was attributed to more mobile electrical charge carriers in the oxide layer on the ZIRLO alloy. The current through the oxide layers on the ZIRLO increased more drastically with increasing temperature than on the NDZ, which indicating that more charge trap sites exist in the ZIRLO than in NDZ. Based on the dynamic charge carrier transport behavior, it was concluded that the electrical charge carrier transport within the oxide layers was closely related with the corrosion behavior of the Zr alloys.

지르칼로이-4의 고온 수증기 산화에서 압력효과

  • 박광헌;김광표;황주호
    • Proceedings of the Korean Institute of Surface Engineering Conference
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    • 2000.05a
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    • pp.5-5
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    • 2000
  • In the severe accident case like LOCA, Zircaloy(Zry) claddings are oxidized not only in high temperature but also in high pressures. It is a concem whether the safety of high bum up fuels can be maintained during severe accident. The effects of steam pressure on Zry-4 oxidation, and the effect of prc-existing oxide layer on the cladding in the high temperature-high pressure oxidation of Ziy-4 were investigated. The experimental temperature range was $700-900^{\circ}C$, and the pressures were between 0.1 and l5.0MPa. Partial pressure of steam tumed out to be the important one rather than total gas pressure. The higher the steam pressure was applied, the thicker the oxide became. nle effect of st,earn pressure on the oxidation of claddings with preexisting oxide was about 40-60% less effective than that of pickled cladding. Aocelerated oxidation in highpressure slean1 seems to be originated from the formation of microcracks produced during the transformation of tetragonal zirconia to monoclinic phase. Steam pressure seems to affect the stability of tetragonal phase.

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Dry storage of spent nuclear fuel and high active waste in Germany-Current situation and technical aspects on inventories integrity for a prolonged storage time

  • Spykman, Gerold
    • Nuclear Engineering and Technology
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    • v.50 no.2
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    • pp.313-317
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    • 2018
  • Licenses for the storage of spent nuclear fuel (SNF) and vitrified highly active waste in casks under dry conditions are limited to 40 years and have to be renewed for prolonged storage periods. If such a license renewal has to be expected since as in accordance with the new site selection procedure a final repository for spent fuel in Germany will not be available before the year 2050. For transport and possible unloading and loading in new casks for final storage, the integrity and the maintenance of the geometry of the cask's inventory is essential because the SNF rod cladding and the cladding of the vitrified highly active waste are stipulated as a barrier in the storage concept. For SNF, the cladding integrity is ensured currently by limiting the hoop stress and hoop strain as well as the maximum temperature to certain values for a 40-year storage period. For a prolonged storage period, other cladding degradation mechanisms such as inner and outer oxide layer formation, hydrogen pick up, irradiation damages in cladding material crystal structure, helium production from alpha decay, and long-term fission gas release may become leading effects driving degradation mechanisms that have to be discussed.

Zricaloy-4 Oxidation Kinetics in High-Pressure High-Temperature Steam and Application to Accident Analysis (고압 고온 수증기에서 지르칼로이-4 산화반응 정량화 및 사고해석에의 응용)

  • 박광헌
    • Journal of Surface Science and Engineering
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    • v.35 no.6
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    • pp.363-370
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    • 2002
  • Empirical equations for the oxide thickness and the weight gain of Zircaloy-4 cladding during the oxidation in high temperature, high pressure steam have been developed. Firstly, the empirical equations for oxide thickness in 1 atm steam in 700~100$0^{\circ}C$ were made, then, the enhancement factor for the steam pressure effects on Zircaloy-4 cladding oxidation in high temperature steam was added. Based on the analysis of the weight fraction of dissolved oxygen in metal layer, empirical equations for the weight gain of Zircaloy-4 in high pressure, high temperature steam were developed. We compare the developed empirical equations with the Baker-Just correlation. The Baker-Just correlation can give a non-conservative estimation of oxidation of Zircaloy-4, depending on the steam pressure. These developed empirical equations can be used for the correct estimation of oxidation of Zircaloy-4 during accident analysis.

Assessing the impact of DIONISIO-SubChanFlow code coupling in nuclear fuel performance simulations

  • Mauricio Exequiel Cazado;Victor Hugo Sanchez-Espinoza;Alejandro Soba
    • Nuclear Engineering and Technology
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    • v.56 no.11
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    • pp.4843-4850
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    • 2024
  • Realistic simulation of nuclear fuel performance requires not only validated models capable of describing the thermomechanical phenomena that take place within the fuel under irradiation conditions, but a detailed description of the thermal hydraulics of the channel surrounding the fuel rods, which provides the boundary conditions of the system. In this work, the main results and outlooks of coupling the thermal hydraulics code SubChanFlow with the fuel performance code DIONISIO are presented. To achieve this, an internal coupling was implemented, wherein DIONISIO is used as a master code controlling SubChanFlow as a thermal hydraulics subroutine replacing the simplified version already embedded in DIONISIO. Several tests were conducted to ensure the performance and quality of the coupling under normal operation conditions as a first approach. In addition, it was observed that the coupling demonstrated a significant improvement in the description of the cladding temperature and related variables, such as oxide thickness and hydrogen uptake, when compared with experimental data.

Destructive Examination of 3 Cycle Burned 14$\times$14 PWR Fuel (삼주기연소 14$\times$14 PWR 핵연료의 핫셀 파괴시험)

  • 이기순;유길성;이영길;민덕기;서항석
    • Nuclear Engineering and Technology
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    • v.21 no.4
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    • pp.332-340
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    • 1989
  • Destructive examination of 14$\times$14 PWR fuel burned for 3 cycles are carried out to investigate the in-reactor fuel performance. The results obtained are as follows; 1) Grain growth is not occured at the fuel center. 2) Fuel density is decreased as the turnup increase, the density is down to 94.4% TD at burnup of 36,000 MWD/MTU. 3) Average thickness of oxide layer on cladding is less than 10 $\mu$m in the lower and middle section, while it is rapidly increased above 20 $\mu$m in the upper section. 4) The rate of hydride production in the cladding is large in the upper section than lower section and is related to the production of oxide on the cladding

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PROPERTIES OF ZR ALLOY CLADDING AFTER SIMULATED LOCA OXIDATION AND WATER QUENCHING

  • Kim, Hyun-Gil;Kim, Il-Hyun;Jung, Yang-Il;Park, Jeong-Yong;Jeong, Yong-Hwan
    • Nuclear Engineering and Technology
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    • v.42 no.2
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    • pp.193-202
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    • 2010
  • In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from $900^{\circ}C$ to $1250^{\circ}C$ and exposed for 300 s, and then cooled to $700^{\circ}C$ before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to $1250^{\circ}C$ for the oxide layer to verify the oxide crystal structure at each oxidation temperature.

High-temperature oxidation behaviors of ZrSi2 and its coating on the surface of Zircaloy-4 tube by laser 3D printing

  • Kim, Jae Joon;Kim, Hyun Gil;Ryu, Ho Jin
    • Nuclear Engineering and Technology
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    • v.52 no.9
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    • pp.2054-2063
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    • 2020
  • The high-temperature oxidation behavior of ZrSi2 used as a coating material for nuclear fuel cladding was investigated for developing accident-tolerant fuel cladding of light water reactors. Bulk ZrSi2 samples were prepared by spark plasma sintering. In situ X-ray diffraction was conducted in air at 900, 1000, and 1100 ℃ for 20 h. The microstructures of the samples before and after oxidation were examined by scanning electron microscopy and transmission electron microscopy. The results showed that the oxide layer of zirconium silicide exhibited a layer-by-layer structure of crystalline ZrO2 and amorphous SiO2, and the high-temperature oxidation resistance was superior to that of Zircaloy-4 owing to the SiO2 layer formed. ZrSi2 was coated on the Zircaloy-4 tube surface using laser 3D printing, and the coated tube was oxidized for 2000 s at 1200 ℃ under a vapor/argon mixture atmosphere. The outer surface of the coated tube was hardly oxidized (10-30 ㎛), while the inner surface of the uncoated tube was significantly oxidized to approximately 300 ㎛.

ESTIMATION OF ALUMINUM AND ARGON ACTIVATION SOURCES IN THE HANARO COOLANT

  • Jun, Byung-Jin;Lee, Byung-Chul;Kim, Myung-Seop
    • Nuclear Engineering and Technology
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    • v.42 no.4
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    • pp.434-441
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    • 2010
  • The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.

Oxidation and Fretting Wear Characteristics of Zirconium Alloy Tubes (지르코늄 합금 튜브의 산화와 프레팅 마멸 특성)

  • Chung, Il-Sup;Lee, Ho-Seong;Lee, Myung-Ho
    • Tribology and Lubricants
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    • v.25 no.4
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    • pp.250-255
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    • 2009
  • Oxidation characteristics of Zirlo and Zircaloy-4 tubes, which are widely used as nuclear power fuel cladding, are studied in steam environment up to $1200^{\circ}C$. Oxidation resistances are compared in terms of the mass increase due to the absorption of oxygen. The evolution of microscopic structure accompanied with the oxidation process is investigated. Also, the influence of oxidation on the fretting wear characteristics of the tubes is studied. Piezo-electrically actuated rig is employed to fret the tubes with cross-contacting arrangement. Wear scar is observed and measured, by using microscopes and a 3D-profiler. The results of fretting wear are quantified in terms of scar size, wear volume and wear coefficient, and compared for the three different tube materials of oxidated Zirlo, virgin Zirlo and Zircaloy-4.