• Nuclear Engineering and Technology
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• 제51권3호
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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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• 제49권4호
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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.

• 에너지공학
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• 제23권4호
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• pp.19-30
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• 2014

지속가능성, 안전성, 핵확산 저항성, 그리고 경제성이 향상된 제4세대 원자로형의 하나로 소듐냉각 고속로가 원자력 선진국을 중심으로 활발히 개발되고 있다. 우리나라가 주도적으로 개발하고 있는 금속연료를 사용하는 소듐냉각고속로는 우수한 피동안전성과 고유안전성을 가지므로 중대사고로의 진전을 조기에 배제할 수 있는 노형으로 평가된다. 또한 소듐냉각고속로는 기존의 사용후핵연료를 재활용하고 자체적으로 재순환 핵주기를 확립함으로써 원자력에너지의 지속성을 향상시킬 수 있다. 이러한 특성으로 인해 많은 나라들이 소듐냉각고속로를 2050년 이전에 도입하는 것을 미래에너지 전략에 포함시키고 있다.

• Nuclear Engineering and Technology
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• 제39권3호
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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.

• Nuclear Engineering and Technology
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• 제50권6호
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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.

• 한국가시화정보학회지
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• 제9권4호
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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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• 제52권4호
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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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• 제54권6호
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• pp.2120-2134
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• 2022

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid-structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian-Eulerian method is used to describe the fluid-structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1071-1082
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• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

• Nuclear Engineering and Technology
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• 제47권3호
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• pp.260-266
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• 2015

Twenty sodium-cooled fast reactors (SFRs) have provided valuable experience in design, licensing, and operation. This paper summarizes the important safety criteria and safety guidelines of intermediate sodium systems, steam generators, decay heat removal systems and associated construction materials and in-service inspection. The safety criteria and guidelines provide a sufficient framework for design and licensing, in particular by new entrants in SFRs.