• 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.47 no.3
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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.

• 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.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 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.660-665
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• 1998

The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scale and verified through the methodology in this paper, which is referred to advanced Liquid Metal Reactor (ALMR). a Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the intergrated code system. Intergrated Code System (ICS) consist of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related are analyzed once-through. Results of conceptual design are attached in this paper.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.6
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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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• v.55 no.4
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• pp.1528-1539
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• 2023

Since the leakage of sodium in an SFR (sodium-cooled fast reactor) causes an explosion upon reaction with air and water, sodium leakages represent an important safety issue. In this study, a novel technique for improving the reliability of sodium leakage detection applying DDVR (dynamic data validation and reconciliation) is proposed and verified to resolve this technical issue. DDVR is an approach that aims to improve the accuracy of a target system in a dynamic state by minimizing random errors, such as from the uncertainty of instruments and the surrounding environment, and by eliminating gross errors, such as instrument failure, miscalibration, or aging, using the spatial redundancy of measurements in a physical model and the reliability information of the instruments. DDVR also makes it possible to estimate the state of unmeasured points. To validate this approach for supporting sodium leakage detection, this study applies experimental data from a sodium leakage detection experiment performed by the Korea Atomic Energy Research Institute. The validation results show that the reliability of sodium leakage detection is improved by cooperation between DDVR and hardware measurements. Based on these findings, technology integrating software and hardware approaches is suggested to improve the reliability of sodium leakage detection by presenting the expected true state of the system.

• Korean Chemical Engineering Research
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• v.57 no.1
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• pp.28-41
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• 2019

We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.