• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2641-2649
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• 2024

Effective management of slightly used nuclear fuel (SUNF) is crucial for both technical and public acceptance reasons. SUNF management, radiotoxicity risk, and associated financial investment and technological capabilities are major concerns in nuclear power production. Reducing the volume of SUNF can simplify its management, and one possible solution is utilizing small modular reactors (SMR) and advanced fuel designs like those with thorium. This research focuses on studying the neutronic performance and radionuclide inventory of three different thorium fuel configurations. The mass of fissile material in thorium-based fuel significantly impacts Kinf, burn-up, and neutron energy spectrum. Compared to uranium, thorium as a fuel produces far fewer transuranic elements and less long-lived fission products (LLFPs) at the end of the core cycle (EOC). However, certain fission product elements produced from thorium-based fuel exhibit higher radioactivity at the beginning of the core cycle (BOC). Physical separation of thorium and uranium in the fuel block, like seed-and-blanket units (SBU) and duplex fuel designs, generate less radioactive waste with lower radioactivity and longer cycle lengths than homogeneous or mixed thorium-uranium fuel. Furthermore, the SBU and duplex feel designs exhibit comparable neutron spectra, leading to negligible differences in SUNF production between the two.

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.211-218
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• 2010

An analysis model on debris transport in the containment floor of pressurized water reactors is developed in which the flow field is calculated by Eulerian conservation equations of mass and momentum and the debris particles are traced by Lagrange equations of motion using the pre-determined flow field data. For the flow field calculation, two-dimensional Shallow Water Equations derived from Navier Stokes equations are solved using the Finite Volume Method, and the Harten-Lax-van Leer scheme is used for accuracy to capture the dry-to-wet interface. For the debris tracing, a simplified two-dimensional Lagrangian particle tracking model including drag force is developed. Advanced schemes to find the positions of particles over the containment floor and to determine the position of particles reflected from the solid wall are implemented. The present model is applied to calculate the transport fraction to the Hold-up Volume Tank in Advanced Power Reactors 1400. By the present model, the debris transport fraction is predicted, and the effect of particle density and particle size on transport is investigated.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.64 no.2
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• pp.254-259
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• 2015

On January 30, 2012 an auxiliary component of Byron Unit 2 was tripped on bus under voltage. The cause of the event was the failure of the C-phase insulator track for the Unit 2 station auxiliary transformer(SAT) revenue metering transformer. In addition to this event, other events have occurred at other plants resulting in an open-phase condition.[1] Therefore, Nuclear Regulatory Commission(NRC) has requested that not only nuclear power plant(NPP) operating company but also its Design Certification(DC) applicant have to prepare open-phase detecting system in their operating plants and design document. In this paper, various open-phase conditions are simulated in NPP using Electromagnetic Transient Program(EMTP) and Atpdraw, and open-phase condition detecting system is proposed for Main Transformer(MT), Unit Auxiliary Transformer(UAT) and SAT connected power line in NPP.

• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• Journal of Electrical Engineering and Technology
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• v.9 no.4
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• pp.1369-1374
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• 2014

Most accidents of medium-voltage cables installed in nuclear power plants result from the initial defect of internal insulators or the initial failure due to poor construction. However, as the service years of plants increase, the possibility of cable accidents is also rapidly increases. This is primarily caused by electric, mechanical, thermal, and radiation stresses. Recently, much attention is paid to the study of cable diagnoses. To date, partial discharge and Tan${\delta}$ measurements are known as reliable methods to diagnose the aging of medium-voltage cables. High frequency partial discharge measurement techniques have been widely used to diagnose cables in transmission and distribution systems. However, the on-line high frequency partial discharge technique has not been used in the nuclear power plants because of the plant shutdown risk, degraded measurement sensitivity, and application problems. In this paper, the partial discharge measurement with a portable device was tried to evaluate the integrity of the 4.16kV and 13.8kV cable lines. The test results show that the high detection sensitivity can be achieved by the high frequency partial discharge technique. The present technique is highly attractive to diagnose medium voltage cables in nuclear power plants.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1279-1288
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• 2019

The vortex-type fluidic diode (FD) is a key safety component for inherent safety in various advanced reactors such as the sodium fast reactor (SFR) and the molten salt reactor (MSR). In this study, topology optimization is conducted to optimize the design of the vortex-type fluidic diode. The optimization domain is simplified to 2-dimensional geometry for a tangential port and chamber. As a result, a design with a circular chamber and a restrictor at the tangential port is obtained. To verify the new design, experimental study and computational fluid dynamics (CFD) analysis were conducted for inlet Reynolds numbers between 2000 and 6000. However, the results show that the performance of the new design is no better than the original reference design. To analyze the cause of this result, detailed analysis is performed on the velocity and pressure field using flow visualization experiments and 3-D CFD analysis. The results show that the discrepancy between the optimization results in 2-D and the experimental results in 3-D originated from exclusion of an important pressure loss contributor in the optimization process. This study also concludes that the junction design of the axial port and chamber offers potential for improvement of fluidic diode performance.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2952-2965
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• 2023

Advanced reactors, such as Small Modular Reactors or existing Nuclear Power Plants, often use Field Programmable Gate Array (FPGA) based controllers in new Instrumentation and Control (I&C) system architectures or as an alternative to existing analog-based I&C systems. Compared to CPU-based Programmable Logic Controllers (PLCs), FPGAs offer better overall performance. However, programming functions on FPGAs can be challenging due to the requirement for a hardware description language that does not explicitly support the operation of real numbers. This study aims to implement the Reactor Trip (RT) functions of the existing analog-based Reactor Protection System (RPS) using FPGAs. The RT equations for Overtemperature delta Temperature and Overpower delta Temperature involve dynamic compensators expressed with the Laplace transform variable, 's', which is not directly supported by FPGAs. To address this issue, the trip equations with the Laplace variable in the continuous-time domain are transformed to the discrete-time domain using the Z-transform. Additionally, a new operation based on a relative value for the equation range is introduced for the handling of real numbers in the RT functions. The proposed approach can be utilized for upgrading the existing analog-based RPS as well as digitalizing control systems in advanced reactor systems.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1574-1583
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• 2024

Small Modular Reactors (SMRs) are compact nuclear reactors designed to generate electric power up to 300 MWe. They could be assembled in factory, and then transported to be directly installed on-stie. CAREM (Central Argentina de Elementos Modulares) is a national SMR development project, based on light water reactor technology supervised by Argentina's National Atomic Energy Commission (CNEA). It is a natural circulation-based SMR with an indirect-cycle, including specific items and parts that simplify the design and improve safety performance. In this paper, the thermal-hydraulic study of CAREM-25 advanced small modular reactor is conducted by using COBRA-EN modified code and the Porous Media Approach (PMA) for the first time. According to PMA approach, each fuel assembly is modeled and divided into a network of lumped regions. While complex geometries are defined, the thermal-hydraulic parameters such as temperature and density are calculated for coolant and fuel rods. The obtained results show that the temperature in the fuel center may reach a peak around 1280 K in the hottest fuel assembly. Finally, the comparison of results from both methods (modified COBRA-EN and PMA) presented an appropriate consistency.

• Applied Chemistry for Engineering
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• v.16 no.2
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• pp.282-287
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• 2005

Palladium membranes, which are permselective to hydrogen separation, were used for the hydrogen purification and in membrane reactors for improving conversions by shifting the reaction equilibrium. Palladium/ceramic composite membranes were prepared by electroless plating technique and then etched in titanium chloride ($TiCl_4$) as a post treatment to enhance the membrane's durability. These membranes were used for membrane reactors in water gas shift (WGS) reaction. CO conversions for the membrane reactor were obtained according to experimental parameters and compared to the traditional reactor without a palladium/ceramic membrane. As a result, CO conversion using palladium membrane reactor at an appropriate condition was over 20~25% greater than that without the membrane reactor. The stability in the long-term test of up to 120 h for WGS reaction with the membrane reactor was good without the degredation of CO conversion.

• Nuclear Engineering and Technology
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• v.37 no.2
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• pp.127-138
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• 2005

Approaches to several recent issues in the operation of nuclear power plants using computational intelligence are discussed. These issues include 1) noise analysis techniques, 2) on-line monitoring and sensor validation, 3) regularization of ill-posed surveillance and diagnostic measurements, 4) transient identification, 5) artificial intelligence-based core monitoring and diagnostic system, 6) continuous efficiency improvement of nuclear power plants, and 7) autonomous anticipatory control and intelligent-agents. Several changes to the focus of Computational Intelligence in Nuclear Engineering have occurred in the past few years. With earlier activities focusing on the development of condition monitoring and diagnostic techniques for current nuclear power plants, recent activities have focused on the implementation of those methods and the development of methods for next generation plants and space reactors. These advanced techniques are expected to become increasingly important as current generation nuclear power plants have their licenses extended to 60 years and next generation reactors are being designed to operate for extended fuel cycles (up to 25 years), with less operator oversight, and especially for nuclear plants operating in severe environments such as space or ice-bound locations.