• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1442-1450
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• 2017

In liquid metal fast breeder reactors, postulated failures of the plant protection system may lead to serious unprotected accidental consequences. Unprotected transients are generically categorized as transient overpower accidents and transient under cooling accidents. In both cases, core meltdown may occur and this can lead to a molten fuel coolant interaction (MFCI). The understanding of MFCI phenomena is essential for study of debris coolability and characteristics during post-accident heat removal. Sodium is used as coolant in liquid metal fast breeder reactors. Viewing inside sodium at elevated temperature is impossible because of its opaqueness. In the present study, a methodology to depict MFCI phenomena using a flat panel detector based imaging system (i.e., real time radiography) is brought out using a woods metal-water experimental facility which simulates the $UO_2-Na$ interaction. The developed imaging system can capture attributes of the MFCI process like jet breakup length, jet front velocity, fragmented particle size, and a profile of the debris bed using digital image processing methods like image filtering, segmentation, and edge detection. This paper describes the MFCI process and developed imaging methodology to capture MFCI attributes which are directly related to the safe aspects of a sodium fast reactor.

• Journal of Energy Engineering
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• v.23 no.4
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• pp.19-30
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• 2014

The leading countries in nuclear technology development are concentrating their efforts on the development of Sodium-cooled Fast Reactor, which is one of the Generation-IV nuclear reactor systems characterized by a sustainability, an enhanced safety, proliferation resistance, and improved economics. Especially, the Republic of Korea is developing a Sodium-cooled Fast Reactor equipped with metallic-fuel. This type of fast reactor has superior inherent safety and passive safety characteristics. Further, sodium-cooled fast reactors enable the reuse of spent fuel and the closing of fuel cycle, thus, it increases the sustainability of nuclear energy. Many countries are planning the deployment of sodium-cooled fast reactors before 2050 in their energy mix.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1110-1119
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• 2020

The Floating Absorber for Safety at Transient (FAST) is a safety device used in the innovative Sodium-cooled Fast Reactor (iSFR). The FAST insert negative reactivity under transient or accident conditions. However, behavior of the FAST is still unclear under transient conditions. Therefore, the existing Floating Absorber for Safety at Transient Analysis Code (FASTAC) is improved to analyze the FAST movement by considering the reactivity and temperature distribution within the reactor core. The current FAST system is simulated under a single control rod withdrawal accident condition. In this investigation, the reactor thermal power does not return to its initial thermal power even if the FAST inserts negative reactivity. Only a 9 K of coolant temperature margin, in the hottest fuel assembly at EOL, can lead to unnecessary insertion of the negative reactivity. On the other hand, the FASTs cannot contribute to controlling the reactivity when normalized radial power is less than 0.889 at BOL and 0.972 at EOL. These simulation results suggest that the current FAST design needs to be optimized depending on its installed location. Meanwhile, the FAST system keeps the fuel, cladding and coolant temperatures below their limit temperatures with given conditions.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.47-54
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• 2010

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/$cm^3$. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1022-1028
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• 2024

A neutronics study of a supercritical CO₂-cooled fast reactor core for nuclear propulsion has been performed in this work. The thermal power of the reactor core is 30 MWth and a ceramic UO₂ fuel can be used to achieve a 20-year lifetime without refueling. In order to make a compact core with inherent safety features, the drum-type reactivity control system and folding-type shutdown system are adopted. In addition, we suggest a cold shutdown system using gadolinium as a spectral shift absorber (SSA) against flooding. Although there is a penalty of U-235 enrichment for the core embedded with the cold shutdown system, it effectively mitigates the increment of reactivity at the flooding of seawater. In this study, the neutronics analyses have been performed by using the continuous energy Monte Carlo Serpent 2 code with the evaluated nuclear data file ENDF/B-VII.1 Library. The supercritical CO₂-cooled fast reactor core is characterized in view of important safety parameters such as the reactivity worth of reactivity control systems, fuel temperature coefficient (FTC), coolant temperature coefficient (CTC), and coolant temperature-density coefficient (CTDC). We can say that the suggested core has inherent safety features and enough flexibility for load-following operation.

• Journal of the Korean Society of Visualization
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• v.9 no.4
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• pp.54-62
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• 2011

KAERI has developed a KALIMER-600 which is a pool-type sodium-cooled fast reactor with a 600MWe electric generation capacity. For a SFR development, one of the main topics is an enhancement of the reactor system safety. Therefore, we have a long-term plan to design the large sodium experimental facility to evaluate the reactor safety and component performance. In order to extrapolate a thermal hydraulic phenomena in a large sodium reactor, the thermal hydraulics phenomena is under investigation in a 1/$10^{th}$ water-simulant facility for the KALIMER-600. In this paper, we shortly described the experimental facility setup and the measurement of the isothermal global flow behavior. For the flow field measurement, the PIV method was used in a transparent Plexiglas reactor vessel model at around $20^{\circ}C$ water condition.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3127-3132
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• 2021

Feasibility study of burn-up analysis and monitoring using delayed fast neutrons was investigated at Missouri University of Science and Technology Reactor (MSTR). Burnt and fresh fuel elements were used to collect delayed fast neutron data for different power levels. Total reactivity varied depending on the burn-up rate of fuel elements for each core configuration. The regulating rod worth was 2.07E-04 𝚫k/k/in and 1.95E-04 𝚫k/k/in for T121 and T122 core configurations at 11 inch, respectively. Delayed fast neutron spectrum of F1 (burnt) and F16 (fresh) fuel elements were analyzed further, and a strong correlation was observed between delayed fast neutron emission and burn-up. According to the analyzed peaks in burnt and fresh fuels, reactor power dependency was observed and it was determined that delayed neutron provided more reliable results at reactor powers of 50 kW and above.

• Clean Technology
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• v.25 no.2
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• pp.161-167
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• 2019

Several types of reactors have been used during the past decade to perform fast pyrolysis of biomass. Among the developed fast pyrolysis reactors, fluidized bed reactors have been widely used in the fast pyrolysis process. In recent years, experimental studies have been conducted on the characteristics of biomass fast pyrolysis in a spouted bed reactor. The fluidization characteristics of a spouted bed reactor are influenced by particle properties, fluid jet velocity, and the structure of the core and annulus. The geometry of the spouted bed reactor is the main factor determining the structure of the core and annulus. Accordingly, to optimize the design of a spouted bed reactor, it is necessary to study the pyrolysis characteristics of biomass. However, no detailed investigations have been made of the fast pyrolysis characteristics of biomass in accordance with the geometry of the spouted bed reactor. In this study, fast pyrolysis experiments using Jatropha curcas L. seed shell cake were conducted in a conical spouted bed reactor to study the effects of reaction temperature and reactor cone angle on the product yield and pyrolysis oil quality. The highest energy yield of pyrolysis oil obtained was 63.9% with a reaction temperature of $450^{\circ}C$ and reactor cone angle of $44^{\circ}$. The results showed that the reaction temperature and reactor cone angle affected the quality of the pyrolysis oil.