• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.993-1001
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• 2024

The TANDEM project is a European initiative funded under the EURATOM program. The project started on September 2022 and has a duration of 36 months. TANDEM stands for Small Modular ReacTor for a European sAfe aNd Decarbonized Energy Mix. Small Modular Reactors (SMRs) can be hybridized with other energy sources, storage systems and energy conversion applications to provide electricity, heat and hydrogen. Hybrid energy systems have the potential to strongly contribute to the energy decarbonization targeting carbon-neutrality in Europe by 2050. However, the integration of nuclear reactors, particularly SMRs, in hybrid energy systems, is a new R&D topic to be investigated. In this context, the TANDEM project aims to develop assessments and tools to facilitate the safe and efficient integration of SMRs into low-carbon hybrid energy systems. An open-source "TANDEM" model library of hybrid system components will be developed in Modelica language which, by coupling, will extend the capabilities of existing tools implemented in the project. The project proposes to specifically address the safety issues of SMRs related to their integration into hybrid energy systems, involving specific interactions between SMRs and the rest of the hybrid systems; new initiating events may have to be considered in the safety approach. TANDEM will study two hybrid systems covering the main trends of the European energy policy and market evolution at 2035's horizon: a district heating network and power supply in a large urban area, and an energy hub serving energy conversion systems, including hydrogen production; the energy hub is inspired from a harbor-like infrastructure. TANDEM will provide assessments on SMR safety, hybrid system operationality and techno-economics. Societal considerations will also be encased by analyzing European citizen engagement in SMR technology safety.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.156-168
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• 2023

This study investigates the high temperature oxidation behaviour of a Ni-20Cr-1.2Si (wt.%) alloy in steam from 1200 ℃ to 1350 ℃ by Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM), Energy Dispersive X-ray Spectroscopy (EDS) and X-ray Diffraction (XRD). The results demonstrate that exposed Ni-based alloy developed a thin oxide scale, consisted mainly of Cr₂O₃. The oxidation kinetics obtained from the experimental results was applied to evaluate the hydrogen generation considering a simplified reactor core model with different cladding alloys following an unmitigated Loss-Of-Coolant Accident (LOCA) scenario in a hypothetical Small Modular Reactor (SMR). Overall, experimental data and simulations results show that both Fe-based and Ni-based alloys may enhance cladding survivability, delaying its melting, as well as reducing hydrogen generation under accident conditions compared to Zr-based alloys. However, a substantial neutron absorption occurs when Ni-based alloys are used as cladding for current uranium-dioxide fuel systems, even when compared to Fe-based alloys.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2534-2544
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• 2024

This paper proposes a flooding safety system (FSS) and its operation strategy that can provide long-term safety and effective maintenance for modules of small modular reactor (SMR) and metal containment maintained at dried environment during normal operation. During hypothesized accidents, the FSS re-collects the evaporated steam into the common pool by the condenser installed above the common water pool and provides an emergency coolant for the cavities and auxiliary pools. This study suggested that the condensate re-collection strategy using the FSS can effectively delay the depletion of available water in response to the accidents. Without recollection, the achievable grace periods ranged from 44 to 1507 days for six-module and one-module accidents, respectively. However, with a full re-collection (ratio = 1.0), the time to total depletion of emergency coolant was estimated indefinite. Even with a partial re-collection ratio of 0.3, a grace period of 83.5 days could be ensured for a six-module transient. This study reported the effectiveness of condensate re-collection and the FSS as an innovative safety management strategy and system. Employing a condensate re-collection strategy with a high re-collection ratio can enhance the long-term safety and effective convenience of SMR operations and maintenance.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1974-1987
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• 2023

The load-following operation of small modular reactors (SMRs) requires accurate prediction of transient behaviors that can occur in the balance of plants (BOP) and the nuclear steam supply system (NSSS). However, 1-D thermal-hydraulics analysis codes developed for safety and performance analysis have conventionally excluded the BOP from the simulation by assuming ideal boundary conditions for the main steam and feed water (MS/FW) systems, i.e., an open loop. In this study, we introduced a lumped model of BOP fluid system and coupled it with NSSS without any ideal boundary conditions, i.e., in a closed loop. Various methods for coupling boundary conditions at MS/FW were tested to validate their combination in terms of minimizing numerical instability, which mainly arises from the coupled boundaries. The method exhibiting the best performance was selected and applied to a transient simulation of an integrated NSSS and BOP system of a SMART. For a transient event with core power change of 100-20-100%, the simulation exhibited numerical stability throughout the system without any significant perturbation of thermal-hydraulic parameters. Thus, the introduced boundary-condition coupling method and BOP fluid system model can expectedly be employed for the transient simulation and performance analysis of SMRs requiring daily load-following operations.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.893-905
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• 2016

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.75-82
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• 2021

The 3D printing technology is one of the fourth industrial revolution technology that drives innovation in the manufacturing process, and should be applied to nuclear industry for various purposes according to the manufacturing trend change. In nuclear industry, it can be applied to manufacture obsolete items and new designed parts in advanced reactors or small modular reactors (SMRs), replacing the traditional manufacturing technologies. A gate valve body was manufactured, which was obsolete in nuclear power plant, using DED(Directed Energy Deposition) metal 3D printing technology after restoring design characteristics including 3D design drawing by reverse engineering. The 3D printed valve body was assembled with commercial parts such as seat-ring, disk, stem, and actuator for performance test. For the valve assembly, including 3D printed valve body, several tests were performed, including pressure test, end-loading test, and seismic test according to KEPIC MGG and KEPIC MFC. In the pressure test, hydraulic pressure of 391kgf/cm² was applied to 3D printed valve body, and no leak was detected. Also the 3D printed valve assembly was performed well in end-loading and seismic tests.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.359-367
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• 2018

Long-lived Small Modular Reactors are being promoted as an innovative way of catering to emerging markets and isolated regions. They can be operated continuously for decades without requiring additional fuel. A novel configuration of long-lived reactor core employs a mixed neutron spectrum, providing an improvement in nonproliferation metrics and in safety characteristics. Starting with a base sodium reactor design, moderating material is inserted in outer core assemblies to modify the fast spectrum. The assemblies are shuffled once during core lifetime to ensure that every fuel rod is exposed to the thermalized spectrum. The Mixed Spectrum Reactor is able to maintain a core lifetime over two decades while ensuring the plutonium it breeds is below the weapon-grade limit at the fuel discharge. The main drawbacks of the design are higher front-end fuel cycle costs and a 58% increase in core volume, although it is alleviated to some extent by a 48% higher power output.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.689-696
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• 2012

A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.