• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.249-260
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• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• Nuclear Engineering and Technology
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• v.40 no.2
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• pp.127-132
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• 2008

An experimental loop operating with water at supercritical conditions(25MPa, $600^{\circ}C$ in the test section) is designed for operation in the research reactor LVR-15 in UJV Rez, Czech Republic. The loop should serve as an experimental facility for corrosion tests of materials for in-core as well as out-of-core structures, for testing and optimization of suitable water chemistry for a future HPLWR and for studies of radiolysis of water at supercritical conditions, which remains the domain where very few experimental data are available. At present, final necessary calculations(thermalhydraulic, neutronic, strength) are being performed on the irradiation channel, which is the most challenging part of the loop. The concept of the primary and auxiliary circuits has been completed. The design of the loop shall be finished in the course of the year 2007 to start the construction, out-of-pile testing to verify proper functioning of all systems and as such to be ready for in-pile tests by the end of the HPLWR Phase 2 European project by the end of 2009.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.31 no.5
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• pp.482-490
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• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests and the simple analysis were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The calculated circulation flow rate was similar to experimental ones within about ${\pm}$15% error bounds and depended on the form loss due to the inlet/outlet area.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.850-857
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• 2006

Hydrogen production using current fueling facilities is essential for near-term applications of fuel cells. A preliminary process for developing a natural gas autothermal reforming (ATR) reactor for fuel cells is presented in this paper. A experimental reactor for methane ATR was constructed and used for characterization of Jin reactor. Temperature profiles of the reactor were observed, and reformed gas compositions were analyzed to evaluate efficiency, conversion and reaction heat with varying amounts of $O_2/CH_4$ at selected furnace temperature and $H_2O/CH_4$. The amount of $O_2/CH_4$ showed strong offsets on reactor temperature, efficiency and conversion indicating that $O_2/CH_4$ is a crucial operation condition. Operation conditions which result in thermal neutrality of ATR reactor system were determined for two cases of an ATR system based on the estimation of enthalpy difference between reactants of assumed inlet temperatures and the products from experimental results. The determined conditions for thermally neutral operations could be used for guidelines to design reformers and for determining the operation parameters of a self sustaining ATR reactor.

• Transactions of the Korean hydrogen and new energy society
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• v.23 no.5
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• pp.545-550
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• 2012

This work is for the design study of natural gas reformer (40 $m^3/hr$ over). We used experimental kinetic data from literature. After that, we set up theoretical model based on experimental reaction kinetic data. The shape of reactor is 1.7 m long and 200 mm dia. with cylinder geometry. Volume of reactor is 53.4 liter. Average flow velocity of gases in the reactor has been determined 0.272 m/sec and residence time is 9.26 sec. Reaction temperature is $850^{\circ}C$, with pressure 9.3 Bar. Used natural gas volume is about 9.21 $m^3/hr$. Produced hydrogen is 43.7 $m^3/hr$ with no change of pressure. Unreacted natural gas is 0.09 $m^3/hr$ and the amount of steam is 26.9 $m^3/hr$. Steam to $CH_4$ (s/c ratio) is 2.91. Reforming reaction take place from the reactor entrance to 120 cm region of cylinder type reactor. After the entrance of reacting gases to 120 cm region, the reaction reaches equilibrium which is close to products. This study can be applicable to design various reactors. Output data is in good agreements with the data in literatures1).

• Journal of computational fluids engineering
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• v.7 no.3
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• pp.44-52
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• 2002

An analytic methodology was developed for free surface motions between liquid metal coolant and cover gas in order to calculate the phenomena of gas entrainment in hot pool surface through IHX EMP and reactor core. The methodology was setup by applying the first order VOF convection model to CFX4 general purpose fluid dynamics analysis code. The methodology was validated by applying it to an experimental apparatus designed for free surface motions of KALIMER reactor. The distributions of free surface calculated by the present methodology were almost coincident with the experimental data. The developed methodology was applied to the KALIMER reactor of full power operating condition. The shapes of the free surface were nearly uniform. From the results, it was found that the altitude of the free surface from the IHX inlet nozzle of KALIMER reactor is high enough not to affect to free surface motions of generating gas bubbles from the turbulent shear flows such as hydraulic jump and water falls.

• Journal of Environmental Science International
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• v.13 no.3
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• pp.319-325
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• 2004

An experimental study on the preparation of monolithic porous polymers by environmentally friend process in supercritical carbon dioxide has been carried out. Polymerization mixture composed of a cross-linking monomer, initiator and functional co-polymer was charged in the reactor with sapphire window. After the system was purged with a flow of $CO_2$ for 15 min, the reactor was pressurized with liquid $CO_2$ up to 100 bars. The reactor was isolated from and placed back to the system via quick connector for shaking until the mixture had become fully homogeneous. The reactor was then heated and pressurized to the required reaction conditions and left overnight. After cooling and $CO_2$ evacuation, the polymer was removed from the reactor as dry, white, continuous monoliths. The effect of experimental conditions on the physical properties of porous polymer was systematically examined, and it was found that monomer content had a major effect on the physical properties of the polymers.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.253-259
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• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2088-2095
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• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.