• Journal of Mechanical Science and Technology
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• v.15 no.7
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• pp.921-930
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• 2001

A new theoretical critical heat flux (CHF) model was developed for the forced convective flow boiling at high pressure, high mass velocity, and low quality. The present model for an intermittent vapor blanket was basically derived from the sublayer dryout theory without including any empirical constant. The vapor blanket velocity was estimated by an axial force balance, and the thickness of vapor blanket was determined by a radial force balance for the Marangoni force and lift force. Based on the comparison of the predicted CHF with the experimental data taken from previous studies, the present CHF model showed satisfactory results with reasonable accuracy.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.503-517
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• 1995

A mechanistic model for forced convective transition boiling has been developed to predict transition boiling heat flux realistically. This model is based on a postulated multi­stage boiling process occurring during the passage time of an elongated vapor blanket specified at a critical heat flux condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling. The total heat transfer rate during the transition boiling is the sum of the heat transfer rates after the DNB weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. From these comparisons, it can be seen that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are nil predicted at low qualities/high pressures near 10 bar.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4274-4281
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• 2023

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1251-1260
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• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.927-938
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• 2023

The blanket plays an important role in fusion reactor and stands extremely high thermal and electromagnetic loads during operation situation and plasma disruption event, brings the need for precise thermal and electromagnetic analysis. Since the thermal field and EM field interact with each other nonlinearly, we develop a method of electromagnetic-thermal two-way coupling by using finite element software COMSOL. The coupling analyses of blanket under steady state and MD event are implemented and the results are analyzed. For steady state, the influences of coupling effects are relatively small but still recommended to be considered for a high precision analysis. The influence of thermal field on EM field can't be ignored under MD events. The variation of force density could cause a significant change in stress in certain parts of blanket. The influence of Joule heat during MD event is negligible, yet the potential temperature rise caused by induced current after MD event still needs to be researched.

• Journal of Environmental Science International
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• v.9 no.2
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• pp.101-108
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• 2000

Through numerical experiment using simplified OSU-1D PBL(Oregon State University One-Dimensional Planetary Boundary Layer) model and field measurement, we studied the impacts of vegetation canopy on heat island that was one of the characteristics of local climaate in urban area. it was found that if the fraction of vegetation was extended by 10 percent, the maximum air temperature and the maximum ground temperature can come down about 0.9${\circ}C$, 2.3${\circ}C$, respectively. Even though the field measurement was done under a little unstable atmospheric condition, the canopy air temperature was lower in the daytime, and higher at night than the air and ground temperature. This result suggests that the extention of vegetation canopy can bring about more pleasant local climate by causing the oasis, the shade and the blanket effect.

• Journal of Veterinary Clinics
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• v.37 no.6
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• pp.331-335
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• 2020

A 13-year-old female Poodle presented with gallbladder rupture caused by a mucocele and hypoperfusion. The animal had a thermal burn injury caused by a forced-air warming device used for hypothermia during surgery. This could be attributed to two causes. First, the forced-air warming device was used with direct hosing, without attachment to an air blanket, and the heat was concentrated in a single area. Second, perioperative peripheral hypoperfusion hampered heat dissipation and increased the susceptibility to a burn injury. These findings suggest that an air blanket should be used with a forced-air warming device according to the manufacturer's instructions. Furthermore, patients with peripheral hypoperfusion are at a higher risk of burn injuries and require close monitoring.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.662-669
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• 2010

To develop the manufacturing methods for the blanket first wall (FW) of the International Thermonuclear Experimental Reactor (ITER) and to verify the integrity of the joint, Be/Cu mockups were fabricated and tested at the KoHLT-1 (Korea Heat Load Test facility), a graphite heater facility located at the Korea Atomic Energy Research Institute (KAERI). Since Be and Cu joining is the focus of the present study, the fabricated mockups had a CuCrZr heat sink joined with three Be tiles as an armor material, unlike the original ITER blanket FW, which has a stainless steel structure and coolant tubes. Hot isostatic pressing (HIP) was carried out at $580^{\circ}C$ and 100 MPa for 2 hours as the method for Be/Cu joining. Three interlayers, namely, $1{\mu}mCr/10{\mu}mCu$, $1{\mu}mTi/0.5{\mu}mCr/10{\mu}mCu$, and $5{\mu}mTi/10{\mu}mCu$ were applied as a coating to the Be tiles by a physical vapor deposition (PVD) method. A shear test was performed with the specimens, which were fabricated by the same methods as those used to fabricate the mockups. The average values were 125 MPa to 180 MPa, and the samples with the $1{\mu}mCr/10{\mu}mCu$ interlayer showed the lowest value. No defect or delamination was found in the joints of the mockups by the developed ultrasonic test using a flat-type probe with a 10 MHz frequency and a 0.25 inch diameter. High heat flux (HHF) tests were performed at $1.0\;MW/m^2$ heat flux for each mockup using the given conditions, and the results were analyzed by ANSYS-CFX code. For the test criteria, an expected fatigue lifetime about 1,000 cycles was obtained by analysis with ANSYS-mechanical code. Mockups using the interlayers of $1{\mu}mTi/0.5{\mu}mCr/10{\mu}mCu$ and $5{\mu}mTi/10{\mu}mCu$ survived up to 1,100 cycles over the required number of cycles. However, one of the Be tiles in the other two mockups using the $1{\mu}mCr/10{\mu}mCu$ interlayer was detached during the screening test, and others were detached by discharge after 862 cycles. The integrity of the joints using the proposed interlayers was proven by the HHF test, but the other interlayer requires more study before it can be used for the joining of Be to Cu. Moreover, it was confirmed that the measured temperatures agreed well with the analysis temperatures, which were used to estimate the lifetime and that the developed facility showed its capability of the long time operation.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.401-422
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• 2005

We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall/blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel structures. The research described for these concepts includes both thermofluid MHD issues for the liquid metal coolant in the DCLL, and thermomechanical issues for ceramic breeder packed pebble beds in the solid breeder concept. Finally, future directions for ongoing research in these areas are described.