• 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.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.248-257
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• 2020

This study presents an initial design for a novel system consisting in a coupled nuclear reactor and a phase change material-based thermal energy storage (TES) component, which acts as a buffer and regulator of heat transfer between the primary and secondary loops. The goal of this concept is to enhance the capacity factor of nuclear power plants (NPPs) in the case of high integration of renewable energy sources into the electric grid. Hence, this system could support in elevating the economics of NPPs in current competitive markets, especially with subsidized solar and wind energy sources, and relatively low oil and gas prices. Furthermore, utilizing a prismatic-core advanced high temperature reactor (PAHTR) cooled by a molten salt with a high melting point, have the potential in increasing the system efficiency due to its high operating temperature, and providing the baseline requirements for coupling other process heat applications. The present research studies the neutronics and thermal hydraulics (TH) of the PAHTR as well as TH calculations for the TES which consists of 300 blocks with a total heat storage capacity of 150 MWd. SERPENT Monte Carlo and MCNP5 codes carried out the neutronics analysis of the PAHTR which is sized to have a 5-year refueling cycle and rated power of 300 MW_th. The PAHTR has 10 metric tons of heavy metal with 19.75 wt% enriched UO₂ TRISO fuel, a hot clean excess reactivity and shutdown margin of $33.70 and -$115.68; respectively, negative temperature feedback coefficients, and an axial flux peaking factor of 1.68. Star-CCM + code predicted the correct convective heat transfer coefficient variations for both the reactor and the storage. TH analysis results show that the flow in the primary loop (in the reactor and TES) remains in the developing mixed convection regime while it reaches a fully developed flow in the secondary loop.

• Journal of Korean Society of Water and Wastewater
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• v.23 no.5
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• pp.539-545
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• 2009

A feasibility of a pilot scale two-phase anaerobic digestion with ultra filtration system treating garbage leachate were evaluated. The treatment system consisted of a thermophilic acidogenic reactor, a mesophilic methanogenic reactor, and an UF membrane. The average COD removal efficiency of the treatment system was 95% up to the OLR of 3.1 g COD/L/d. The higher COD removal efficiency with membrane unit resulted from the removal of some portion of soluble organics by membrane as well as particulate materials. When the membrane unit was in operation, bulk liquid in acidogenic and methanogenic reactors was partially interchanged, which maintained the acidogenic reactor pH over 5.0 without external chemical addition. Also, with the production of methane in the acidogenic reactor, the organic loading rate of the methanogenic reactor reduced. The initial flux of the membrane unit was $50{\sim}60L/m^2/hr$, but decreased to $5 L/m^2/hr$ after 95 days of operation due to clogging caused by particulate materials such as fibrous materials in garbage leachate. To prevent clogging caused by particulate materials, a pretreatment system such as screening is required. With the improvement with membrane unit operation, the two-phase anaerobic digestion with ultra filtration system is expected to have the possibility of treating garbage leachate.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.581-594
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• 2014

A feasibility study on the seismic design of nuclear reactor buildings with application of a seismic isolation system is introduced. After the Hyogo-ken Nanbu earthquake in Japan of 1995, seismic isolation technologies have been widely employed for commercial buildings. Having become a mature technology, seismic isolation systems can be applied to NPP facilities in areas of high seismicity. Two reactor buildings are discussed, representing the PWR and BWR buildings in Japan, and the application of seismic isolation systems is discussed. The isolation system employing rubber bearings with a lead plug positioned (LRB) is examined. Through a series of seismic response analyses using the so-named standard design earthquake motions covering the design basis earthquake motions obtained for NPP sites in Japan, the responses of the seismic isolated reactor buildings are evaluated. It is revealed that for the building structures examined herein: (1) the responses of both isolated buildings and isolating LRBs fulfill the specified design criteria; (2) the responses obtained for the isolating LRBs first reach the ultimate condition when intensity of motion is 2.0 to 2.5 times as large as that of the design-basis; and (3) the responses of isolated reactor building fall below the range of the prescribed criteria.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.338-344
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• 2010

A double rotation C-scanning system and a software program Under-Sodium MultiVIEW have been developed for the under-sodium viewing of a reactor core and in-vessel structures of a sodium-cooled fast reactor KALIMER-600. Double rotation C-scanning system has been designed and manufactured by the reproduction of double rotation plug of a reactor head in KALIMER-600. Hardware system which consists of a double rotating scanner, ultrasonic waveguide sensors, a high power ultrasonic pulser-receiver, a scanner driving module and a multi channel A/D board have been constructed. The functions of scanner control, image mapping and signal processing of Under-Sodium MultiVIEW program have been implemented by using a LabVIEW graphical programming language. The performance of Under-Sodium MultiVIEW program was verified by a double rotation C-scanning test in water.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.922-932
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• 2024

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl₃ fuel and NaCl-KCl-MgCl₂ coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 1993.04a
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• pp.159-164
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• 1993

This paper presents a computer design program for In-Core Instrumentation(ICI) guide tube routing and locations on support system, and checking the interference between ICI guide tubes in the reactor coolant system of typical Pressurized Water Reactor. The program, ICITRIC, has been written in FORTRAN language which is available under UNIX environment. Results of this program are compared with those of the commercial code, PATRAN, and both results are almost same Also the results may provide input data for ICI system static and dynamic analysis performed by the commercial code, SUPER PIPE. This program can simulate ICI guide tube routing and locations on support system, and checking the interference between ICI guide tubes. Through a process of iteration, the designer can apply initial conditions, and modify the routing until satisfied with the overall system performance.

• Journal of Korean Society on Water Environment
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• v.21 no.3
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• pp.256-262
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• 2005

The purpose of this study is to investigate the performances of organic removal and methane recovery by using a full scale two-phase anaerobic system. The full scale two-phase anaerobic process was consists of an acidogenic anaerobic baffled reactor (ABR) and a methanognic upflow anaerobic sludge blanket (UASB) reactor. The volumes of acidogenic and methanogenic reactors were designed to $28.3m^3$ and $75.3m^3$. The two-phase anaerobic system represented 60-82% of COD removal efficiency when the influent COD concentration was in the range of 7,150 to 16,270 mg/L after screening (average concentration is 10,280 mg/L). After steady-state, the effluent COD concentration in the methanogenic reactor showed $2,740{\pm}330 mg/L$ by representing average COD removal efficiency was $71.4{\pm}8.1%$ when the operating temperature was in the range of $19-32^{\circ}C$. The effluent SCOD concentration was in the range of 2,000-3,000 mg/L at the steady state while the volatile fatty acid concentration was not detected in the effluent. Meanwhile, the COD removal efficiency in the acidogenic reactor showed less than 5%. The acidogenic reactor played key roles to reduce a shock-loading when periodic shock loading was applied and to acidify influent organics. Due to the high concentration of alkalinity and high pH in the effluent of the methanogenic reactor, over 80% of methane in the biogas was produced consistently. More than 70% of methane was recovered from theoretical methane production of TCOD removed in this research. The produced gas can be directly used as a heat source to increase the reactor temperature.

• Journal of the Korean Institute of Gas
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• v.20 no.6
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• pp.95-101
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• 2016

For a natural gas steam reforming, comparative studies of the performance in a conventional packed-bed reactor and a membrane reactor, a new conceptual reactor consisting of a reactor with series of hydrogen separation membranes, have been performed. Based on experimental kinetics reported by Xu and Froment, a process simulation model was developed with Aspen $HYSYS^{(R)}$, a commercial process simulator, and effects of various operating conditions like temperature, $H_2$ permeance, and Ar sweep gas flow rate on the performance in a membrane reactor were investigated in terms of reactant conversion and $H_2$ yield enhancement showing improved $H_2$ yield and methane conversion in a membrane reactor. In addition, a preliminary cost estimation focusing on natural gas consumption to supply heat required for the system was carried out and feasibility of possible cost savings in a membrane reactor was assessed with a cost saving of 10.94% in a membrane reactor.