• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.171-171
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• 2001

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.411-416
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• 2006

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the comer 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.329-338
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• 2010

We developed an achievable control logic for the reactor power level during a power maneuvering event and set up some constraints for the control of the reactor power in a conceptual sodium-cooled fast reactor (KALIMER-600) that was developed at KAERI. For simulating the dynamic behaviors of the plant, we developed a fast-running performance analysis code. Through various simulations of the power maneuvering event, we evaluated some suggested control logic for the reactor power and found an achievable control logic. The objective of the control logic is to search for the position of the control rods that would keep the average temperature of the primary pool constant and, concurrently, minimize the power deviation between the reactor and the BOP cycle during the power maneuvering. In addition, the flow rates of the primary pool and the intermediate loop should be changed according to the power level in order to not violate the constraints set up in this study. Also, we evaluated some movement speeds of the control rods and found that a fast movement of the control rods might cause the power to fluctuate during the power maneuvering event. We suggested a reasonable movement speed of the control rods for the developed control logic.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.353-363
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• 1998

The H$_{\infty}$ algorithms of the mixed weight sensitivity is used for the robust design of the reactor power control system. The mixed weight sensitivity method requires the selection of the proper weighting functions for the loop shaping in frequency domain. The complexity of the system equation and the non-convexity of the problem make it very difficult to determine the weighting functions. The genetic algorithm which is improved and hybridized with the simulated annealing is applied to determine the weighting functions. This approach permits an automatic calculation and the resultant system shows good robustness and performance.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.167-174
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• 1997

The method that the operating condition of Control Rod Drive Mechanism (CRDM) can be monitored without mounting sensors within CRDM housing was developed, and by using this developed method the closed-loop controller for the CRDM was designed which can optimize the performance and maximize the reliability of CRDM operation. Neural network is utilized as pattern recognition engine in detecting CRDM actuation. In this paper, most problems in previous open loop system are resolved. The control algorithms for closed-loop system ore developed and implemented within the hardware of timing controller based on microprocessor. All functions in the timing controller ore verified by means of real time CRDM simulator. The results show that the timing controller performs its intended functions properly.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3597-3611
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• 2021

The present paper studies the thermal-hydraulic behaviour of the rectangular supercritical natural circulation loop (SCNCL) using numerical model of one dimensional. Then the results of this model is confirmed with experimental and benchmark results. Variations with several geometric parameters like loop diameter, riser length, and heater length and operating conditions like heater inlet enthalpy, pressure, friction factor, and inlet and exit loss coefficient on steady-state performance are investigated for various orientations like HHHC, HHVC, VHVC and VHHC of the heater and cooler. The chances of existing static instability (Ledinegg excursion) has been investigated, which reveals that it can arise only in a low inlet enthalpy condition, far from the suggested various orientations of heater and cooler.

• Journal of the Korean Mathematical Society
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• v.49 no.3
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• pp.641-658
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• 2012

The goal of this paper is to develop a nonlinear observer-based control strategy for a multi-variables continuous stirred tank reactor (CSTR). A new robust nonlinear observer is constructed to estimate the whole process state variables. The observer is coupled with a nonlinear controller, designed based on the input-output linearization for controlling the concentration and reactor temperature. The closed loop system is shown to be globally asymptotically stable based on Lyapunov arguments. Finally, computer simulations are developed for showing the performance of the proposed controller.

• Journal of Applied Reliability
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• v.10 no.1
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• pp.57-64
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• 2010

The HANARO nuclear research reactor had been operated successfully and smoothly up to the year 2008 since its first year of instability in year 1995 just after the completion of construction. But the reliability of the reactor has been degraded from the year 2009 due to new experimental facilities such as Feul Test Loop(FTL) and Cold Neutron Source(CNS) which were installed in the HANARO plant. It turned out that these new facilities contributed unexpected stoppage of the plant. This paper describes causes of stoppage and suggestions to improve the reliability of the plant.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.552-559
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• 2020

This paper presents an observer-based chattering free robust optimal control scheme to regulate the total power of a nuclear reactor. The non-linear model of nuclear reactor is linearized around a steady state operating point to obtain a linear model for which an optimal second order integral sliding mode controller is designed. A second order integral sliding mode observer is also designed to estimate the unmeasurable states. In order to avoid the chattering effect, the discontinuous input of both observer and controller are designed using the super-twisting algorithm. The proposed controller is realized by combining an optimal linear tracking controller with a second order integral sliding mode controller to ensure minimum control effort and robustness of the closed-loop system in the presence of uncertainties. The condition for the selection of gains of discontinuous control based on the super-twisting algorithm is derived using a strict Lyapunov function. Performance of the proposed observer based control scheme is demonstrated through non-linear simulation studies.