• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3207-3216
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• 2021

Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partially blocked 7-pin fuel assembly mock-up of sodium-cooled fast reactor have been investigated through a numerical analysis using a commercial computational fluid dynamics code, ANSYS CFX. The simulation with the SST turbulence model agrees well with the experimental data of outlet and cladding wall temperatures. From the analysis on the limiting streamline at the wall, multi-scale vortexes developed in axial direction were found around the blockage. The vortex core has a high cladding wall temperature, and the attachment line has a low cladding wall temperature. The small-scale vortex structures significantly enhance the convective heat transfer because it increases the turbulent mixing and the turbulence kinetic energy. The large-scale vortex structures supply thermal energy near the heated cladding wall surface. It is expected that control of the vortex structures in the fuel assembly plays a significant role in the convective heat transfer enhancement. Furthermore, the blockage plate and grid spacer increase the pressure drop to about 36% compared to the bare case.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.932-941
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• 2021

This paper suggests a fracture simulation method for SFR metallic fuel pin under accident condition. Two major failure mechanisms - creep damage and eutectic penetration - are implemented in the suggested method. To simulate damaged element, stress-reduction concept to reduce stiffness of the damaged element is applied. Using the proposed method, the failure size of cladding can be predicted in addition to the failure time and failure site. To verify the suggested method, Whole-pin furnace (WPF) test and TREAT-M test conducted at Argonne National Laboratory (ANL) are simulated. In all cases, predicted results and experimental results are overall in good agreement. Based on the simulation result, the effect of eutectic-penetration depth representing failure behavior on failure size is studied.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.772-784
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• 2024

The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to use line and surface integrals of polynomials in a deformed hexagonal geometry. The nodal calculation is accelerated by the coarse mesh finite difference (CMFD) formulation extended to unstructured geometry. The accuracy of the unstructured nodal solution was evaluated for a group of 2D SFR core problems in which the assembly corner points are arbitrarily displaced. The RENUS results for the change in nuclear characteristics resulting from fuel deformation were compared with those of the reference McCARD Monte Carlo code. It turned out that the two solutions agree within 18 pcm in reactivity change and 0.46% in assembly power distribution change. These results demonstrate that the proposed unstructured nodal method can accurately model heterogeneous thermal expansion in hexagonal fueled cores.

• Korean Chemical Engineering Research
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• v.54 no.6
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• pp.863-870
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• 2016

We investigated the physical/chemical phenomena that a slow loss of $CO_2$ inventory into sodium after the sodium-$CO_2$ boundary failure in printed circuit heat exchangers (PCHEs), which is considered for the supercritical $CO_2$ Brayton cycle power conversion system of a sodium-cooled fast reactor (SFR). The first phenomenon is plugging inside narrow sodium channels by micro cracks and the other one is damage propagation referred to as wastage combined with the corrosion/erosion effect. Experimental results of plugging shows that sodium flow immediately stopped as $CO_2$ was injected through the nozzle at $300{\sim}400^{\circ}C$ in 3 mmID sodium channels, whereas sodium flow stopped about 60 min after $CO_2$ injection in 5 mmID sodium channels. These results imply that if pressure boundary of sodium-$CO_2$ fails a narrow sodium channel would be plugged by reaction products in a short time whereas a relatively wider sodium channel would be plugged with higher concentration of reaction products. Wastage by the erosion effect of $CO_2$ (200~250 bar) hardly occurred regardless of the kinds of materials (stainless steel 316, Inconel 600, and 9Cr-1Mo steel), temperature ($400{\sim}500^{\circ}C$), or the diameter of the $CO_2$ nozzle (0.2~0.8 mm). Velocities at the $CO_2$ nozzle were specified as Mach 0.4~0.7. Our experimental results are expected to be used for determining the design parameters of PCHEs for their safeties.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.77-84
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• 2010

Reduction of oxides has been investigated for the volume reduction and recycling of the spent oxide fuel from commercial nuclear power plants. Various oxide reduction methods were proposed and KAERI (Korea Atomic Energy Research Institute) is currently developing an electrochemical reduction process using a LiCl-$Li_2O$ molten salt as a reaction medium. The electrochemical reduction process, the front end of the pyroprocessing, can connect the PWR (Pressurized Water Reactor) oxide fuel cycle to a metal fuel cycle of the sodium cooled fast reactor. This paper summarizes KAERI efforts on the development, improvement, and scale-up of the oxide reduction process.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.12
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• pp.1421-1426
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• 2014

To improve the inherent safety of the sodium-cooled fast reactor (SFR), the supercritical $CO_2$ ($S-CO_2$) Brayton cycle is being considered as an alternative power conversion system to steam the Rankine cycle. In the $S-CO_2$ system, a PCHE (printed circuit heat exchanger) is being considered. In this type of heat exchangers, diffusion bonding is used for joining the thin plates. In this study, the diffusion bonding characteristics of various austenitic alloys were evaluated. The tensile properties were measured at temperatures starting from the room temperature up to $650^{\circ}C$. For the 316H and 347H types of stainless steel, the tensile ductility was well maintained up to $550^{\circ}C$. However, the Incoloy 800HT showed lower strength and ductility at all temperatures. The microstructure near the bond line was examined to understand the reason for the loss of ductility at high temperatures.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.12
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• pp.1415-1420
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• 2014

Super-critical $CO_2$ ($S-CO_2$) Brayton cycle has been considered to replace the current steam Rankine cycle in Sodium-cooled Fast Reactor (SFR) in order to improve the inherent safety and thermal efficiency. Several austenitic alloys are considered as the structural materials for high temperature $S-CO_2$ environment.. Microstructural change after long-term exposure to high temperature $S-CO_2$ environment could affect to the mechanical properties. In this study, candidate materials (austenitic stainless steels and Alloy 800HT) were exposed to $S-CO_2$ to assess oxidation resistance and the change in tensile properties. Loss of ductility was observed for some austenitic stainless steels even after 250 h exposure. The contribution of $S-CO_2$ environment on such changes was analyzed based on the characterization of the surface oxide and carburization of the materials in which 316H and 800H showed different oxidation behaviors.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.9
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• pp.1083-1090
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• 2011

A steam generator poses many difficulties during the development of a sodium-cooled fast reactor because of the sodium-water-reaction problems. Until now, several types of steam generators have been developed, but the specifications of these generators differed in each country. Moreover, even if a country had developed a steam generator, it was not used in the subsequent reactor because the current techniques were not stabilized to select the proper steam generator. As a common development, the Benson steam cycle with few welding locations and high economical efficiency may be adopted. Moreover, the design is dwelled on the convenience of inspection, detection, control, and maintenance for the wear caused by sodiumwater reactions. The specifications of the designed steam generators were reviewed and the current technical problems for steam generators were analyzed. Concepts were proposed to overcome the current technical problems for steam generators.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1611-1625
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• 2020

We propose a scaled-down experimental model of vertical air-natural convection channels by applying the modified Ishii-Kataoka scaling method with the assistance of numerical analyses to the Reactor Vault Cooling System (RVCS) of the Proto-type Gen-IV Sodium-cooled fast reactor (PGSFR) being developed in Korea. Two major non-dimensional numbers (modified Richardson and Friction number) from the momentum equation and Stanton number from the energy balance equation were identified to design the scaled-down experimental model to assimilate thermal-hydraulic behaviors of the natural convective air-cooling channel of RVCS. The ratios of the design parameters in the PGSFR RVCS between the prototype and the scaled-down model were determined by setting Richardson and Stanton number to be unity. The friction number which cannot be determined by the Ishii-Kataoka method was estimated by numerical analyses using the MARS-KS system code. The numerical analyses showed that the friction number with the form loss coefficient of 2.0 in the scale-down model would result in an acceptable prediction of the thermal-hydraulic behavior in RVCS. We also performed experimental benchmarking using the scaled-down model with the MARS-KS simulations to verify the appropriateness of the scale-down model, which demonstrated that the temperature rises and the average air flow velocity measured in the scale-down model.