• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1834-1845
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• 2021

Liquid lead-bismuth cooled fast reactor is one of the most promising reactor types among the fourth-generation nuclear energy systems. The flow and heat transfer characteristics of lead-bismuth eutectic (LBE) are completely different from ordinary fluids due to its special thermal properties, causing that the traditional Reynolds analogy is no longer recommended and appropriate. More accurate turbulence flow and heat transfer model for the liquid metal lead-bismuth should be developed and applied in CFD simulation. In this paper, a specific CFD solver for simulating the flow and heat transfer of liquid lead-bismuth based on the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model was developed based on the open source platform OpenFOAM. Then the advantage of proposed model was demonstrated and validated against a set of experimental data. Finally, the simulation of LBE turbulent flow and heat transfer in a 7-pin wire-wrapped rod bundle with the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model was carried out. The influence of wire on the flow and heat transfer characteristics and the three-dimensional distribution of key thermal hydraulic parameters such as temperature, cross-flow velocity and Nusselt number were studied and presented. Compared with the traditional SED model with a constant Pr_t = 1.5 or 2.0, the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model is more accurate on predicting the turbulence flow and heat transfer of liquid lead-bismuth. The average relative error of the k - 𝜀 - k_𝜃 - 𝜀_𝜃 model is reduced by 11.1% at most under the simulation conditions in this paper. This work is meaningful for the thermal hydraulic analysis and structure design of fuel assembly in the liquid lead-bismuth cooled fast reactor.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.257-264
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• 2015

To maintain sustainability of nuclear energy as an important energy source, both safety problem and Spent Nuclear Fuels(SNFs) problem should be solved. In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be a suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) can be possibly used as a coolant to use fast neutrons. In this paper, it was described that natural circulation parameter studies, design analyses, material selections and a completion of facilities. To develop a natural circulation facility, thermal hydraulic analyses were performed. Installation technique of liquid metal natural circulation were secured.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.293-302
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• 2019

Many researches have been done to develop and improve the performance of personal (individual) dosimeter response to cover a wide of neutron energy range (from thermal to fast). Depending on the individual category of the dosimeter, the semiconductor sensor has been used to simplify and lightweight. In this plan, it's very important to have a fairly accurate counting of doses rate in different energies. With a general design and single-sensor simulations, all optimal thicknesses have been extracted. The performance of the simulation scheme has been compared with the commercial and laboratory samples in the world. Due to the deviation of all dosimeters with a flat energy response, in this paper, has been used an idea of one semi-conductor sensor to have the flat energy-response in the entire neutron energy range. Finally, by analyzing of the sensors data as arrays for the first time, we have reached a nearly flat and acceptable energy-response. Also a comparison has been made between Lucite-PMMA ($H_5C_5O_2$) and polyethylene-PE ($CH_2$) as a radiator and $B_4C$ has been studied as absorbent. Moreover, in this paper, the effect of gamma dose in the dosimeter has been investigated and shown around the standard has not been exceeded.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.942-948
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• 2021

The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.359-374
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• 2024

A high-fidelity numerical analysis methodology was proposed for evaluating the fuel rod cladding integrity of a Prototype Gen IV Sodium Fast Reactor (PGSFR) during normal operation and Design basis events (DBEs). The MARS-LMR code, system transient safety analysis code, was applied to analyze the DBEs. The results of the MARS-LMR code were used as boundary condition for a 3D computational fluid dynamics (CFD) analysis. The peak temperatures considering HCFs satisfied the cladding temperature limit. The temperature and pressure distributions were calculated by ANSYS CFX code, and applied to structural analysis. Structural analysis was performed using ANSYS Mechanical code. The seismic reactivity insertion SSE accident among DBEs had the highest peak cladding temperature and the maximum stress, as the value of 87 MPa. The fuel cladding had over 40 % safety margin, and the strain was below the strain limit. Deformation behavior was elucidated for providing relative coordinate data on each active fuel rod center. Bending deformation resulted in a flower shape, and bowing bundle did not interact with the duct of fuel assemblies. Fuel rod maximum expansion was generated with highest stress. Therefore, it was concluded that the fuel rod cladding of the PGSFR has sufficient structural safety margin during DBEs.

• Journal of Energy Engineering
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• v.12 no.4
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• pp.289-301
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• 2003

To develop a chemical-looping combustion technology, conceptual design of 50 kW thermal chemical-looping combustor, which is composed of two interconnected pressurized circulating fluidized beds, was performed by means of mass and energy balance calculations. A riser type fast fluidized bed was selected as an oxidizer and a bubbling fluidized bed was selected as a reducer by mass balance for the chemical-looping combustor. Calculated values of bed mass, solid circulation flux, and reactor dimension by mass and energy balance calculations were suitable for construction and operation of chemical-looping combustor. It is concluded from the comparison of the design results and operating values of commercial circulating fluidized bed that the process outline is realistic. Moreover, the previous results support that oxygen carrier particle, NiO/bentonite, fulfills the conversion rates needed for the proposed design. The effects of system capacity, metal oxide content in a oxygen carrier particle, amount of steam input, gas velocity, and solid depth on design values were investigated and the changes in the system performance can be estimated by proposed design tool.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• 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.

• Corrosion Science and Technology
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• v.20 no.5
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• pp.308-314
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• 2021

Neutron dose level at bottom head of a reactor pressure vessel (RPV) was calculated using reactor vessel neutron transport for a Korean nuclear power plant A. At 34 EFPY with a 40-year (2042) design life after plating repair, irradiation fast neutron effect was 6.6x10¹⁵ n/cm². As helium(He) gas can be generated by Ni only at 1/10⁶ level of 5 × 10²¹ n/cm², He generation possibility in the Ni plating layer is very little during 40 years of operation (2042, 34 EFPY). Thermal neutrons can significantly affect the generation of He from Ni metal. At 10 years after a repair, He can be generated at a level of about 0.06 appm, a level that can add general welding repair without any consideration. After 40 years of repair, 9.8 appm of He may be generated. Although this is a rather high value, it is within the range of 0.1 to 10 appm when welding repair can be applied. Clad repair by Ni electroplating technology is expected to greatly improve the operation efficiency by improving the safety and shortening the maintenance period of the nuclear power plant.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1557-1564
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• 2020

In this paper, 3D model-based interactive gamma ray shielding package (MIGSHIELD) is developed in virtual reality platform for windows operating system. In MIGSHIELD, the computational methodology is based on point kernel algorithm (PK), several key parameters of PK are obtained using new technique and new methods. MIGSHIELD has interactive capability with virtual world. The main features made in the MIGSHIELD are (i) handling of physical information from virtual world, (ii) handling of arbitrary shapes radioactive source, (iii) calculating the mean free path of gamma ray, (iv) providing interactive function between PK and virtual world, (v) making better use of PK for virtual simulation, (vi) plug and play. The developed package will be of immense use for calculations involving radiation dose assessment in nuclear safety and contributing to fast radiation simulation for virtual nuclear facilities.