• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.922-929
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• 2019

This paper presents the drop performance test of the control rod assembly which is one of the main components strongly related to the safety of the prototype generation IV sodium-cooled fast reactor. To investigate the drop performance, a real-sized control rod assembly that was recently modified based on the drop analysis results was newly fabricated, and several free drop tests under different flow rate conditions were carried out. Then the results were compared with those obtained from the previous tests conducted on the conceptually designed control rod assembly to demonstrate the improvement in performance. Moreover, the drop performance tests under several types and magnitudes of seismic loadings were also conducted to investigate the effect of the seismic loading on the drop performance of the modified control rod assembly. The results showed that the effects of the type and magnitude of the seismic loading on the drop performance of the modified control rod assembly were not significant. Also, the drop time requirement was successfully satisfied, even under the seismic loading conditions.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.855-864
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• 2017

The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.187-192
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• 2007

The basic research on fast reactor technology was started in the mid-1960's in China. The emphasis was put on fast reactor neutronics, thermohydraulics, sodium technology, materials, fuels, safety, sodium devices and instrumentation. In 1987, the research turned to applied basic research with the conceptual design of a 60 MW experimental fast reactor as a target. The Project of the China Experimental Fast Reactor(CEFR) with a thermal power 65 MW was launched in 1993. The R&D of fast reactor technology then carried out to serve a design demonstration connected with the different phases of the conceptual, preliminary and detailed design of the CEFR. Recently, three directions of fast rector technology R&D activities have been considered, and some research programs have been developed. They are: (1) R&D related to the CEFR, i.e. experiments to be conducted on the CEFR for its safe operation, (2) R&D related to the projects of a prototype and the demonstration of fast reactors, and(3) advanced SFR technology within the framework of the international cooperation of INPRO and GIF.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.171-186
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• 2007

Critical issues in the development targets for the future fast reactor(FR) cycle system, including sodium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast reactor Cycle Technology development(FaCT) Project launched in 2006.

• 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.50 no.6
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• pp.849-853
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• 2018

Mechanistic calculations for wet scrubbing of aerosol/vapor from gas bubble rising in liquid pool are essential to safety of sodium-cooled fast reactor. Hence, scrubbing of volatile fission product from mixed gas bubble rising in sodium pool is presented in this study. To understand this phenomenon, a theoretical model has been setup based on classical theories of aerosol/vapor removal from bubble rising through liquid pools. The model simulates pool scrubbing of sodium iodide aerosol and cesium vapor from a rising mixed gas bubble containing xenon as the inert species. The scrubbing of aerosol and vapor are modeled based on deposition mechanisms and Fick's law of diffusion, respectively. Studies were performed to determine the effect of various key parameters on wet scrubbing. It is observed that for higher vapor diffusion coefficient in gas bubble, the scrubbing efficiency is higher. For aerosols, the cut-off size above which the scrubbing efficiency becomes significant was also determined. The study evaluates the retention capability of liquid sodium used in sodium-cooled fast reactor for its safe 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.

• Journal of the Korean Society for Nondestructive Testing
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• v.23 no.5
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• pp.520-525
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• 2003

Fast Breeder Test Reactor (FBTR) is a 40MW (thermal) / 13.2MW (electrical), Plutonium - Uranium mixed carbide fuelled, sodium cooled, loop type nuclear reactor operating at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. Its main aim is to generate experience in operation of fast reactors and sodium systems and to serve as an irradiation facility for development of fuels and structural materials fur fast reactors. Nuclear reactors pose difficulties to the NDT techniques used to monitor the conditions of the internal components. Sodium cooled fast breeder reactors have their own typical difficulties in using the NDT techniques. These are due to the need for operation in aggressive environment of nuclear radiation and sodium (molten/vapour), as well as the need to maintain leak tightness of a very high order during all states of reactor operation and shutdown for fuel handling, maintenance and remote inspection. This paper discusses the following NDT techniques, which have been successfully used for the past 15 years in FBTR: (i) Periscope and Projector, (ii) Core Co-ordinate Measuring Device and, (iii) Optical fiberscope. The inspection using these techniques have given confidence for further reactor operation at high power by giving useful data on the conditions of the components inside the reactor vessel.

• 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.

• 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.