• 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.37 no.6
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• pp.565-574
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• 2005

The CANFLEX Mk-V bundle is designed to improve upon the critical heat flux (CHF) characteristics of the CANFLEX Mk-IV bundle. The main difference between these two bundles is an increase in bearing pad height of about 0.3 mm in the CANFLEX Mk-IV bundle. This change in bearing pad height leads to an increase in gap flow at the bottom of the bundle, primarily eliminating the localized narrow-gap effect that limits the CHF of the CANFLEX Mk-IV bundle. The objective of this paper is to examine the effects of bearing pad height and pressure tube creep on the sheath-temperature distribution, dryout power, and dryout location, as observed ken full-scale bundle tests, between CANFLEX Mk-IV and Mk-V bundles In uncrept and crept channels. A comparison of surface-temperature differences between the top and bottom elements of the bundles showed that increasing the bearing pad height has led to a more homogeneous enthalpy distribution in subchannels of the bundle. Initial dryout locations of the CANFLEX Mk-V bundle were mainly observed at the mid-spacer plane of either the $10^{th}$ (about $80\%$) or $11^{th}$ ($20\%$) bundle in the 12-bundle string, as compared to the mid-spacer and downstream-button planes for the CANFLEX Mk-IV bundle. Dryout power and boiling-length-average (BLA) CHF values exhibit consistent trends and little scatter with varying flow conditions for both types of CANFLEX bundles in uncrept and crept channels. An increase in pressure tube creep has led to a reduction in dryout power (about $20\%$ far the $3.3\%$ crept channel and $27\%$ for the $5.1\%$ crept channel as compared to dryout powers for the uncrept channel). Increasing the bearing pad height of the CANFLEX bundle has led to an increase in the dryout power. Overall, the dryout power of the CANFLEX Mk-V bundle is 7 to $10\%$ higher than that of the CANFLEX Mk-IV bundle at the inlet temperature range of interest (i.e., between 243 and $290^{\circ}C$).

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2755-2770
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• 2022

The need for Dissimilar Welded Joint (DWJ) in the power plant components arises in order to increase the overall efficiency of the plant and to avoid premature failure in the component welds. The Activated-Tungsten Inert Gas (A-TIG) welding process, which is a variant of Tungsten Inert Gas (TIG) welding, is focus of this review work concerning the DWJ of nuclear grade creep-strength enhanced ferritic/martensitic (CSEF/M) steels and austenitic steels. A-TIG DWJs are compared with Multipass-Tungsten Inert Gas (M-TIG) DWJ based on their mechanical and microstructural properties. The limitations of multipass welding have put A-TIG welding in focus as A-TIG provides a weld with increased depth of penetration (DOP) and enhanced mechanical properties. Hence, this review article covers the A-TIG welding principle and working parameters along with detailed analysis of role played by the flux in welding procedure. Further, weld characteristics of martensitic and austenitic steel DWJ developed with the A-TIG welding process and the M-TIG welding process are compared in this study as there are differences in mechanical, microstructural, creep-related, and residual stress obtained in both TIG variants. The mechanics involved in the welding process is deliberated which is revealed by microstructural changes and behavior of base metals and WFZ.

• Progress in Superconductivity and Cryogenics
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• v.25 no.3
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• pp.28-33
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• 2023

Magnetic relaxation properties of pure MgB₂ powder samples and diluted water-treated MgB₂ powder samples were investigated. The magnetic field H-dependence, m(H), and the time t-dependence, m(t), of the magnetic moment m were measured and analyzed using the PPMS-VSM magnetometer equipment, respectively. The m(t) reduction rates of pure MgB₂ powder samples and diluted water-treated MgB₂ powder samples decreased to about 0.7 ~ 1.8% and 0.6 ~ 1.0% for about 7200 s, respectively, at temperature T = 15 K. The magnetic relaxation properties of the two types of MgB₂ powders were analyzed by calculating the magnetic relaxation rate S = -dln(M_irr)/dln(t) values according to Anderson-Kim theory. The magnetic relaxation ratio S values of the two types of MgB₂ powder samples were almost similar. As a result of the quantum creep effect, the constant magnetic relaxation rate S characteristic was confirmed at a temperature range of T = 10 K or less.

• Journal of Magnetics
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• v.3 no.3
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• pp.82-85
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• 1998

A portable HTS RF SQUID-based system, weighing less than 20 kg has been built for susceptometry applications in weak magnetic fields, It includes a YBCO sensor for measuring the axial magnetic field component with a resolution of about $7{\times}10^{-13} T/Hz^{1/2}.$ This is determined by the intrinsic magnetic noise in the quasi-white noise region. There is a relaxation for a sudden increase in field due to magnetic flux creep in HTS. In this instance the time did not exceed 3~5 minutes.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2454-2465
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• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.27-35
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• 2023

Pressure tubes are the main components of PHWR core and serve as the pressure boundary of the primary heat transport system. However, because pressure tubes have changed their geometrical dimensions under the severe operating conditions of high temperature, high pressure and neutron irradiation according to the increase of operation time, all dimensional changes should be predicted to ensure that dimensions remain within the allowable design ranges during the operation. Among the deformations, the diameter expansion due to creep leads to the increase of bypass flow which may not contribute to the fuel cooling, the decrease of critical channel power and finally the deration of the power to maintain the operational safety margin. This study is focused on the modeling of the expansion of the pressure tube diameter based on the operating conditions and measured diameter data. The pressure tube diameter expansion was modeled using the neutron flux and temperature distributions of each fuel channel and each fuel bundle as well as the measured diameter data. Although the basic concept of the current modeling approach is simple, the diameter prediction results using the developed methodology showed very good agreement with the real data, compared to the existing methodology.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.423-430
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• 2014

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

• Journal of the Microelectronics and Packaging Society
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• v.21 no.4
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• pp.83-89
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• 2014

Solder paste plays a crucial role as the widely used joining material in surface mount technology (SMT). The understanding of its behaviour and properties is essential to ensure the proper functioning of the electronic assemblies. The composition of the solder paste is known to be directly related to its rheological behaviour. This paper provides a brief overview of the solder paste behaviour of four different solder paste formulations, stencil printing processes, and techniques to characterize solder paste behaviour adequately. The solder pastes are based on the Sn-3.0Ag-0.5Cu alloy, are different in their particle size, metal content and flux system. The solder pastes are characterized in terms of solder particle size and shape as well as the rheological characterizations such as oscillatory sweep tests, viscosity, and creep recovery behaviour of pastes.

• Journal of Welding and Joining
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• v.30 no.6
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• pp.10-14
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• 2012

Recently the low alloy steel plate made with manganese-molybdenum is used widely in steam drum and separator of the new coal-fired power plant boiler. This material is suitable for the vapor storage of high pressure and high temperature. The high temperature creep strength of Mn-Mo alloy is higher than the carbon plate(SA516) that used in the subcritical pressure boiler. It reduces the thickness of the pressure vessel and makes the lightweight possible. Recently in the power plant boiler operation and production process, the damage has happened frequently in the heat affected zone and base material according to the hydrogen crack and delayed crack. This paper describes the research result about the damage case experienced in the boiler steam drum production process and present the optimum manufacture method for the similar damage prevention of recurrence.