• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1135-1144
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• 2024

With the increasing ratio of renewables in the grid, a low-carbon and stable base load source that also is capable of load tracking is in demand. Sodium cooled fast reactors (SFRs) coupled to thermal heat storage system (THS) is a strong candidate for the need. This research focuses on the designing and performance validation of a two-tank THS based on molten salt to integrate with a 280 MW_th sodium cooled fast reactor. Designing of the THS includes the vital component, sodium-to-salt heat exchanger which is a technology gap that needs to be filled, and designing and parameter selection of the tanks and related pumps. Modeling of the designed THS is conducted followed by the description of operation strategies and control logics of the THS. Finally, the dynamic simulation of the designed THS is conducted based on Fortran. Results show, the proposed power system meets the need of the design requirements to store heat for 18 h during a day and provide 500 MW_th for peak demand for the rest of the day.

• 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.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.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.330-338
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• 2016

The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1874-1889
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• 2022

In this study, we evaluate the thermal-hydraulic performance and economics of Printed Circuit Heat Exchanger (PCHE) according to the channel types and associated shape variables for the design of recuperators with Sodium-cooled Fast Reactors (SFRs). To perform the evaluations with variables such as the Reynolds number, channel types, tube diameter, and shape variables, a code for the heat exchanger is developed and verified through a comparison with experimental results. Based on the code, the volume and pressure drop are calculated, and an economic assessment is conducted. The zigzag type, which has bending angle of 80° and a tube diameter of 1.9 mm, is the most economical channel type in a SFR using CO₂ as the working fluid. For a SFR using N₂, we recommend the airfoil type with vertical and horizontal numbers of 1.6 and 1.1, respectively. The airfoil type is superior when the mass flow rate is large because the operating cost changes significantly. When the mass flow rate is small, volume is a more important design parameter, therefore, the zigzag type is suitable. In addition, we conduct a sensitivity analysis based on the production cost of the PCHE to identify changes in optimal channel types.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.902-907
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• 2023

To understand the eutectic reaction mechanism and the relocation behavior of the core debris is indispensable for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (B₄C) and stainless-steel (SS). The influence of the existence of carbon on the B₄C-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the B₄C-SS samples, a new layer was formed between B₄C/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or Fe₂B. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples. All samples observed the Cr-rich domain and Fe and Ni-rich domain after the reaction. These domains might be formed during the solidifying process.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2047-2052
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• 2023

Risk-informed approach has been widely applied in the safety design, regulation, and operation of nuclear reactors. It has been commonly accepted that risk-informed design optimization should be used in the innovative reactor designs to make nuclear system highly safe and reliable. In spite of the risk-informed approach has been used in some advanced nuclear reactors designs, such as Westinghouse IRIS, Gen-IV sodium fast reactors and lead-based fast reactors, the process of risk-informed design of nuclear reactors is hardly to carry out when passive system reliability should be integrated in the framework. A practical method for new passive safety reactors based on probabilistic safety assessment (PSA) and passive system reliability analyze linking is proposed in this paper. New three-dimension frequency-consequence curve based on risk concept with three variables is used in this method. The proposed method has been applied to the determination optimization of design options selection in a 10 MWth lead-based research reactor(LR) to obtain one optimized system design in conceptual design stage, using the integrated reliability and probabilistic safety assessment program RiskA, and the computation resources and time consumption in this process was demonstrated reasonable and acceptable.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.565-578
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• 2022

Among Generation IV reactors, the sodium-cooled fast reactor (SFR) is attracting attention as a system having great potential for commercial use. Gas entrainment is a thermal-hydraulic issue related to the safety problem of the reactor core in the SFR. Typically, a dipped plate or baffles are installed under the free surface to suppress gas entrainment. However, these approaches can cause gas entrainment in other locations and require many trial-and-error and verifications. In this study, a new strategy using magnetohydrodynamics to suppress gas entrainment in the SFR is proposed. In a counter-flow model, a judgment criterion of gas entrainment occurrence was developed for both water and liquid metal. Moreover, the gas entrainment can be completely suppressed by applying a magnetic field.