• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.04a
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• pp.743-744
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• 2012

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.700-711
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• 2015

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called "self-wastage phenomena." A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodiumwater reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.21-26
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• 2012

Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at $1170^{\circ}C$ after the induction melting to make round bar as 160mm diameter, 7000mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2~3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120mm.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1380-1385
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• 2020

An annular linear induction electromagnetic pump (ALIP) which has a developed pressure of 0.76 bar and a flow rate of 100 L/min is designed to analysis end effect which is main problem to use ALIP in thermohydraulic system of the prototype generation-IV sodium-cooled fast reactor (PGSFR). Because there is no moving part which is directly in contact with the liquid, such as the impeller of a mechanical pump, an ALIP is one of the best options for transporting sodium, considering the high temperature and reactivity of liquid sodium. For the analysis of an ALIP, some of the most important characteristics are the electromagnetic properties such as the magnetic field, current density, and the Lorentz force. These electromagnetic properties not only affect the performance of an ALIP, but they additionally influence the end effect. The end effect is caused by distortion to the electromagnetic field at both ends of an ALIP, influencing both the flow stability and developed pressure. The electromagnetic field distribution in an ALIP is analyzed in this study by solving Maxwell's equations and using numerical analysis.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2733-2742
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• 2020

This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P₀ scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P₅, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC²-3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to -223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in ²³⁷Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for ²³⁵U, from 2.2% to 0.7% for ²³⁸Pu, and from 1.6% to 0.2% for ²⁴¹Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important.

• 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.52 no.8
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• pp.1677-1688
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• 2020

In the sodium-cooled fast reactor, the steam generator is a heat exchange device between sodium and water, which may cause leakage, resulting in a sodium-water reaction accident, which in turn affects the safe operation of the entire nuclear reactor. To this end, the electromagnetic vortex flowmeter is used to detect leakage of the steam generator and its signal processing method is studied in this paper. The hydraulic experiment was carried out by using water instead of liquid sodium, and the sensor output signal of the electromagnetic vortex flowmeter under different gas injection volumes was collected. The bubble noise signal is reflected by the base line of the sensor output signal. According to the relationship between the proportion of the bubble noise signal in the sensor output signal and the gas injection volume, a signal processing method based on the energy ratio calculation is proposed to detect whether the water contains bubbles. The gas injection experiment of liquid sodium was conducted to verify the effectiveness of the signal processing method in the detection of bubbles in sodium, and the minimum detectable leak rate of water in the steam generator was detected to be 0.2 g/s.

• 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.52 no.3
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• pp.589-596
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• 2020

Many components in the assembly section of Sodium-cooled Fast Reactor are made of good corrosionresistant 316 LN Stainless Steel material. To avoid self-welding of the components with the coolant sodium at elevated temperature, hardfacing is inevitable. Ni-based colmonoy-5 is used for hardfacing due to its lower dose rate by Plasma Transferred Arc process due to its low dilution. Since Ni-Cr-B-Si alloy becomes very fluidic while depositing, the major height of the weld overlay rests inside the groove. Hardfacing is also done over the plain surface where grooving is not possible. Therefore, grooved and ungrooved hardfaced specimens were prepared at different travel speeds. Fe content at every 100 ㎛ of the weld overlay was studied by Energy Dispersive Spectroscopy and also the micro hardness was determined at those locations. A correlation between iron dilution from the base metal and the micro hardness was established. Therefore, if the Fe content of the weld overlay is known, the hardness at that location can be obtained using the correlation and vice-versa. A new correlation between micro hardness and dilution coefficient is obtained at different locations. A comparative study between those specimens is carried out to recommend the optimum travel speed for lower dilution.