• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.528-533
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• 1997

A CANFLEX(CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5㎜) to reduce maximum element power rating and buttons to enhance the critical heat flux(CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant H-l34a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power(CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF's for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only lot high pressure conditions. It indicates that the button is a useful design lot CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.411-416
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• 2006

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the comer 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2359-2366
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• 2022

Selecting graphite grades with superior irradiation characteristics is important task for designers of graphite moderation reactors. To provide reference information and data for graphite selection, the effects of irradiation on three fine-grained, iso-molded nuclear grade graphites, ETU-10, IG-110, and NBG-25, were compared based on irradiation-induced changes in volume, thermal conductivity, dynamic Young's modulus, and coefficient of thermal expansion. Data employed in this study were obtained from reported irradiation test results in the high flux isotope reactor (HFIR)(ORNL) (ETU-10, IG-110) and high flux reactor (HFR)(NRL) (IG-110, NBG-25). Comparisons were made based on the irradiation dose and irradiation temperature. Overall, the three grades showed similar irradiation-induced property change behaviors, which followed the historic data. More or less grade-sensitive behaviors were observed for the changes in volume and thermal conductivity, and, in contrast, grade-insensitive behaviors were observed for dynamic Young's modulus and coefficient of thermal expansion changes. The ETU-10 of the smallest grain size appeared to show a relatively smaller VC to IG-110 and NBG-25. Drastic decrease in the difference in thermal conductivity was observed for ETU-10 and IG-110 after irradiation. The similar irradiation-induced properties changing behaviors observed in this study especially in the DYM and CTE may be attributed to the assumed similar microstructures that evolved from the similar size coke particles and the same forming method.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.386-391
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• 1997

It is important to understand correctly a CHF under low flow condition for the purpose of enhancing the reactor safety and performance in the LWRs. The CHF experiments have been carried out for an internally heated vertical annulus in RCS loop facility. The experimental conditions cover ranges of pressure from 1.82 to 12.08 MPa, mass flux from 300 to 550kg/$m^2$. s and inlet subcooling of 210kJ/kg. The CHF data decrease with increasing pressure at high value of mass flux. For mass flux of about 300kg/$m^2$. s, the CHF rue little influenced by pressure. The CHF data are correlated well by using the dimensionless heat flux and dimensionless mass flux for a fixed inlet subcooling except the data group of 12.08 MPa. It seems that the Doerffer correlation and Katto correlation overestimate the CHF for low pressure and lower value of mass flux within this experimental ranges. The Bowling correlation gives a better prediction than the other two correlations.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• Journal of Advanced Marine Engineering and Technology
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• v.10 no.2
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• pp.156-164
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• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2139-2146
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• 2023

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 ℃ and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ~268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m² for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.327-336
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• 2020

The high-flux advanced neutron application reactor (HANARO) is a research reactor with thermal power of 30 MW applied in various research and development using neutrons generated from uranium fission chain reaction. A degasifier tank is installed in the ancillary facility of HANARO. This facility generates gas pollutants produced owing to internal environmental factors. The degasifier tank is designed to maintain the gas contaminants below acceptable levels and is monitored using an analyzer in the gas sampling panel. If condensate water is generated and flows into the analyzer of the gas sampling panel, corrosion occurs inside the analyzer's measurement chamber, which causes failure. Condensate water is generated because of the temperature difference between the degasifier tank and analyzer when the gas flows into the analyzer. A heating system is installed between the degasifier tank and gas sampling panel to suppress condensate water generation and effectively remove the condensate water inside the system. In this study, we investigated the efficiency of the heating system. In addition, the variations in the pipe temperature and the amount of average condensate water were modeled using a wall condensation model based on the changes in the fluid inlet temperature, outside air temperature, and heating cable-setting temperature.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2005.11a
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• pp.167-168
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• 2005

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-Tc superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2006.12a
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• pp.112-113
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• 2006

The operational characteristics of the integrated three-phase flux-lock type superconducting fault current limiter (SFCL) were analyzed. The suggested three-phase SFCL consisted of a three-phase flux-lock reactor and three high-$T_c$ superconducting (HTSC) elements. The former has three windings wound on an iron core, each of which has the same turn's ratio between coil 1 and coil 2. The latter are connected in series with coil 2 of each phase. The integrated three-phase flux-lock type SFCL showed the operational characteristics that the fault phase could affect the sound phase, which resulted in quenching the HTSC element in the sound phase. Through the computer simulation applying numerical analysis for its three-phase equivalent circuit, the fault current limiting characteristics of the integrated three-phase flux-lock type SFCL according to the ground fault types were compared.