• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.191-202
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• 2013

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1323-1332
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• 2009

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

• Journal of computational fluids engineering
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• v.19 no.4
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• pp.29-36
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• 2014

A computational study of a thermal striping in the upper plenum of PGSFR(Prototype Generation-IV Sodium-cooled Fast Reactor) being developed at the KAERI(Korea Atomic Energy Research Institute) is presented. The LES(Large Eddy Simulation) approach is employed for the simulation of thermal striping in the upper plenum of the PGSFR. The LES is performed using the WALE (Wall-Adapting Local Eddy-viscosity) model. More than 19.7 million unstructured elements are generated in upper plenum region of the PGSFR using the CFX-Mesh commercial code. The time-averaged velocity components and temperature field in the complicated upper plenum of the PGSFR are presented. The time history of temperature fluctuation at the eight locations of solid walls of UIS(Upper Internal Structure) and IHX(Intermediate Heat eXchanger) are additionally stored. It has been confirmed that the most vulnerable regions to thermal striping are the first plate of UIS. From the temporal variation of temperature at the solid walls, it was possible to find the locations where the thermal stress is large and need to assess whether the solid structures can endure the thermal stress during the reactor life time.

• Journal of computational fluids engineering
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• v.17 no.4
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• pp.41-48
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• 2012

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1869-1873
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• 2008

In this work, to make it possible to generate glow discharge in atmospheric pressure condition with relatively high and wide electric field, micro channel reactor is proposed. Si DRIE and Cr deposition by Ebeam evaporation is used to make channel and bottom electrode layer. Upper electrode is made from ITO glass to visualize discharge within micro channel. Fabricated reactor is verified by generating uniform glow plasma with N2 / He gases each as working fluid. The range of gas electric field to generate glow plasma is from about 200 V/cm and upper limit is not observed in tested condition of up to 150 kV/cm. This data shows that micro channel plasma reactor is more versatile. Indirect estimation of electron temperature in this reactor can be inferred that the electron temperature within glow discharge in micro reactor lies $0{\sim}2eV$. This research demonstrates that the reactor is appropriate in application that needs to maintain low temperature condition during chemical process.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.776-783
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• 2020

As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.43-50
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• 2007

A partial flow blockage in an assembly of a liquid metal reactor could result in a cooling deficiency of the core. To develop a partial blockage detection system, we have studied the changes of the temperature fluctuation characteristics in the upper plenum according to changes of the t10w blockage conditions in an assembly. We analyzed the temperature fluctuation in the upper plenum with the Large Eddy Simulation (LES) turbulence model in the CFX code and evaluated its statistical parameters. Based on the results of the statistical analyses, we developed a neural network model for detecting a partial flow blockage in an assembly. The neural network model can retrieve the size and the location of a flow blockage in an assembly from a change of the root mean square, the standard deviation, and the skewness in the temperature fluctuation data. The neural network model was found to be a possible alternative by which to identify a flow blockage in an assembly of a liquid metal reactor through learning and validating various flow blockage conditions.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.