• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2556-2563
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• 2021

Metaheuristic algorithms can work well in solving or optimizing problems, especially those that require approximation or do not have a good analytical solution. Particle swarm optimization (PSO) is one of these algorithms. The response quality of these algorithms depends on the objective function and its regulated parameters. The nonlinear nature of the pressurized light-water nuclear reactor (PWR) dynamics is a significant target for PSO. The two-point kinetics model of this type of reactor is used because of fission products properties. The proportional-integral-derivative (PID) controller is intended to control the power level of the PWR at a short-time transient. The absolute error (IAE), integral of square error (ISE), integral of time-absolute error (ITAE), and integral of time-square error (ITSE) objective functions have been used as performance indexes to tune the PID gains with PSO. The optimization results with each of them are evaluated with the number of function evaluations (NFE). All performance indexes achieve good results with differences in the rate of over/under-shoot or convergence rate of the cost function, in the desired time domain.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.715-731
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• 2021

Uncertainties are calculated for pressurized water reactor (PWR) spent nuclear fuel (SNF) characteristics. The deterministic code STREAM is currently being used as an SNF analysis tool to obtain isotopic inventory, radioactivity, decay heat, neutron and gamma source strengths. The SNF analysis capability of STREAM was recently validated. However, the uncertainty analysis is yet to be conducted. To estimate the uncertainty due to nuclear data, STREAM is used to perturb nuclear cross section (XS) and resonance integral (RI) libraries produced by NJOY99. The perturbation of XS and RI involves the stochastic sampling of ENDF/B-VII.1 covariance data. To estimate the uncertainty due to modeling parameters (fuel design and irradiation history), surrogate models are built based on polynomial chaos expansion (PCE) and variance-based sensitivity indices (i.e., Sobol' indices) are employed to perform global sensitivity analysis (GSA). The calculation results indicate that uncertainty of SNF due to modeling parameters are also very important and as a result can contribute significantly to the difference of uncertainties due to nuclear data and modeling parameters. In addition, the surrogate model offers a computationally efficient approach with significantly reduced computation time, to accurately evaluate uncertainties of SNF integral characteristics.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.940-947
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• 2022

Meta-heuristic algorithms have found their place in optimization problems. Henry gas solubility optimization (HGSO) is one of the newest population-based algorithms. This algorithm is inspired by Henry's law of physics. To evaluate the performance of a new algorithm, it must be used in various problems. On the other hand, the optimization of the proportional-integral-derivative (PID) gains for load-following of a nuclear power plant (NPP) is a good challenge to assess the performance of HGSO. Accordingly, the power control of a pressurized water reactor (PWR) is targeted, based on the point kinetics model with six groups of delayed-neutron precursors. In any optimization problem based on meta-heuristic algorithms, an efficient objective function is required. Therefore, the integral of the time-weighted square error (ITSE) performance index is utilized as the objective (cost) function of HGSO, which is constrained by a stability criterion in steady-state operations. A Lyapunov approach guarantees this stability. The results show that this method provides superior results compared to an empirically tuned PID controller with the least error. It also achieves good accuracy compared to an established GA-tuned PID controller.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2522-2534
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• 2020

It is well known that the model of nuclear reactors features natural nonlinearity, and variable parameters during power tracking operation. In this paper, a disturbance observer-based adaptive sliding mode control (DOB-ASMC) strategy is proposed for power tracking of the pressurized-water reactor (PWR) in the presence of lumped disturbances. The nuclear reactor model is firstly established based on point-reactor kinetics equations with six delayed neutron groups. Then, a new sliding mode disturbance observer is designed to estimate the lumped disturbance, and its stability is discussed. On the basis of the developed DOB, an adaptive sliding mode control scheme is proposed, which is a combination of backstepping technique and integral sliding mode control approach. In addition, an adaptive law is introduced to enhance the robustness of a PWR with disturbances. The asymptotic stability of the overall control system is verified by Lyapunov stability theory. Simulation results are provided to demonstrate that the proposed DOB-ASMC strategy has better power tracking performance than conventional sliding mode controller and PID control method as well as conventional backstepping controller.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.420-424
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• 1994

This paper presents an improved water level control scheme for Pressurized Water Reactor(PWR) Steam Generator(S/G) at the low power operation and transient states. To reduce fluctuations of the water level by the swell and shrink phenomena, the scheme adds feedforward terms considering S/G pressure and the feedwater temperature into the conventional proportional-integral feedback controller. The simulation results using the Compact Nuclear Simulator show that smaller level errors and much faster settling time than those of the conventional scheme can be obtained. The proposed algorithm is easily implementable and has a potential for the real applications.

• The Transactions of the Korean Institute of Electrical Engineers
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• v.34 no.12
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• pp.491-497
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• 1985

An applicaton of the linear optimal control theory to the problem or load follow control in pressurized water reactors (PWR) is investigated. In order to perform the steady-state and load follow operation in PWR, a nonlinear model for the reactor and steam generator is derived and linearized at 50% rated power. Simulation tests are performed for 10% demanded load. Comparing the dynamic response of the newly developed optimal load follow controller with those of the integral error feedback controller proposed by Yang, the rise time of dynamic response of the former is about 15 seconds faster than those of the latter, thus the results indicate that the fast response of the optimal load follow controller is verified. The results of this work are directly applicable to the design of the load follow control systems for commercially operated PWRs.