• The Journal of Korean Institute of Communications and Information Sciences
• /
• v.30 no.3B
• /
• pp.138-143
• /
• 2005

This paper aims to optimize an access probe algorithm for the CDMA 2000 system. The incremental value of PWR_STEP increases as NUM_STEP as 1dBm±0.2 for access probe of the area with good receiving sensitivity when the mobile station transmits via access channel and does not receive any ACK message. However, for the area with weak receiving sensitivity, according to the algorithm of open-loop power control, the transmitting power amplifier becomes saturated and PWR_STEP incremental value keeps performing access probe to 0dBm±0.2. Therefore interference and battery consumption increases according to the transmission of the mobile station. We have optimized the access probe algorithm according to the receiving sensitivity. We transmit the incremental value of access probe power, with delaying as much RT slot value as indicated by IS-95C standard in case of good receiving sensitivity and with delaying RT+l slot value in case of weak receiving sensitivity. Simulation results showed that the proposed algorithm contributes to decrease of the interference and battery consumption according to the transmitting power of the mobile station and improves the call duration.

• Nuclear Engineering and Technology
• /
• v.51 no.2
• /
• pp.356-368
• /
• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2048-2054
• /
• 2022

The coolant temperature feedback coefficient is an important parameter of reactor core power control system. To study the coolant temperature feedback coefficient influence on the core power control system of small PWR, the core power control system is built with the nonlinear model and fuzzy control theory. Then, the uncertainty quantification method of reactor core parameters is established based on the Latin hypercube sampling method and the Bootstrap method. Finally, under the conditions of reactivity step perturbation and coolant inlet temperature step perturbation, uncertainty analysis for two cases is carried out. The result shows that with fuzzy controller and fuzzy PID controller, the uncertainty of the coolant temperature feedback coefficient affects the core power control system, and the maximum uncertainties of core relative power, coolant temperature deviation, fuel temperature deviation and total reactivity are acceptable.

• Nuclear Engineering and Technology
• /
• v.51 no.5
• /
• pp.1195-1208
• /
• 2019

This paper introduces a new two-step procedure for PWR depletion analyses. This procedure adopts the albedo-corrected parameterized equivalence constants (APEC) method to correct the lattice-based raw cross sections (XSs) and discontinuity factors (DFs) by accounting for neutron leakage. The intrinsic limitations of the conventional two-step methods are discussed by analyzing a 2-dimensional SMR with the commercial DeCART2D/MASTER code system. For a full-scope development of the APEC correction, the MASTER nodal code was modified so that the group constants can be corrected in the middle of a microscopic core depletion. The basic APEC methodology is described and color-set problems are defined to determine the APEC functions for burnup-dependent XS and DF corrections. Then the new two-step method was applied to depletion analyses of the SMR without thermal feedback, and its validity was evaluated in terms of being able to predict accurately the reactor eigenvalue and nodal power profile. In addition, four variants of the original SMR core were also analyzed for a further evaluation of the APEC-assisted depletion. In this work, several combinations of the burnup-dependent and -independent XS and DF corrections were also considered. The results show that the APEC method could enhance the nodal equivalence significantly with inexpensive additional costs.

• Journal of the Korea Institute of Information and Communication Engineering
• /
• v.3 no.4
• /
• pp.765-778
• /
• 1999

본 논문에서는 CDMA 이동통신 시스템의 모델을 설정하교 역방향 전력제어에 대한 모델을 제안하였으며 전력제어 파라메터에 따른 CDMA 이동통신 시스템의 성능을 평가하였다. 이때 역방향 전력제어 성능 평가 파라메터는 역방향 링크 전력 제어 주기 Tp, 전력제어 지연시간 kTp 그리고 콤맨드 에러 CMD_ERR와 전력의 증가 및 감소량을 나타내는 PWR_STEP등이며, 이들 파라메터들의 평가 결과는 Tp에 대하여 단말기의 이동 속도가 빠를수록 수신 신호 전력 레벨은 기준 레벨에서 심하게 변하는 잔류 페이딩을 볼 수 있었다. 콤맨드 에러에 의한 영향보다는 전력제어 시간 지연에 의한 영향이 더 크게 나옴을 볼 수 있었으며 PWR_STEP의 변화에 따른 영향은 변화량이 약 2dB로 커질 때 전력 제어 오차가 작아짐을 보였다. 이러한 결과 고찰을 통하여 CDMA 이동통신 시스템의 역방향 전력제어 시뮬레이션을 통해 전력제어 파라미터를 최적화함으로써 CDMA이동통신 시스템의 성능을 개선할 수 있음을 보였다.

• Nuclear Engineering and Technology
• /
• v.15 no.2
• /
• pp.98-109
• /
• 1983

A two-step alternating direction explicit method is developed to solve the space-dependent reactor kinetics equations in two space dimensions. As a special case in the general class of alternating direction implicit methods, this method is analysed for accuracy and stability. To test the validity of this method it is compared with the implicit-difference method used in the TWIGL program. It is shown that the two methods are closely related. The time dependent neutron fluxes of the pressurized water reactor (PWR), during control rod insertion, and, of the CANDU-PHW reactor, in case of postulated loss of coolant accident, are obtained from the numerical calculation results.

• Nuclear Engineering and Technology
• /
• v.38 no.1
• /
• pp.81-92
• /
• 2006

In this study, the core dynamics of a PWR reactor is identified online by a recursive least-squares method. Based on the identified reactor model consisting of the control rod position and the core average coolant temperature, the future average coolant temperature is predicted. A model predictive control method is applied to designing an automatic controller for the thermal power control of PWR reactors. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, this procedure for solving the optimization problem is repeated. The objectives of the proposed model predictive controller are to minimize both the difference between the predicted core coolant temperature and the desired temperature, as well as minimizing the variation of the control rod positions. In addition, the objectives are subject to the maximum and minimum control rod positions as well as the maximum control rod speed. Therefore, a genetic algorithm that is appropriate for the accomplishment of multiple objectives is utilized in order to optimize the model predictive controller. A three-dimensional nuclear reactor analysis code, MASTER that was developed by the Korea Atomic Energy Research Institute (KAERI) , is used to verify the proposed controller for a nuclear reactor. From the results of a numerical simulation that was carried out in order to verify the performance of the proposed controller with a $5\%/min$ ramp increase or decrease of a desired load and a $10\%$ step increase or decrease (which were design requirements), it was found that the nuclear power level controlled by the proposed controller could track the desired power level very well.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.70-75
• /
• 1997

A CASMO/ORIGEN coupling utility program was developed to predict the composition of all the fission products in spent PWR fuels. The coupling program reads the CASMO output file, modifies the ORIGEN cross section library and reconstructs the ORIGEN input file at each depletion step. In ORIGEN, the burnup equation is solved for actinides and fission products based on the fission reaction rates and depletion flux of CASMO. A sample calculation has been performed using a 14$\times$14 PWR fuel assembly and the results are given in this paper.

• Proceedings of the KIEE Conference
• /
• 1993.07a
• /
• pp.261-264
• /
• 1993

This paper proposes an adaptive generalized predictive control with feedforward algorithm for steam generator level control in nuclear power plant. The proposed algorithm is shown that the parameters of N-step ahead predictors can be obtained using the parameters of one-step ahead predictor which is derived from plant model with feedforward. Using this property the proposed scheme is an adaptive algorithm which consists of GPC method and the recursive least squares algorithm for identifying the parameters of one-step ahead predictor. Also, computer simulations are performed to evaluate the performance of proposed algorithm using a mathematical model of PWR steam generator Simulation results show good performances for load variation. And the proposed algorithm shows better responses than PI controller does.

• Nuclear Engineering and Technology
• /
• v.49 no.6
• /
• pp.1135-1142
• /
• 2017

The applicability of the algebraic collapsing acceleration (ACA) technique to the method of characteristics (MOC) in cases with scattering anisotropy and/or linear sources was investigated. Previously, the ACA was proven successful in cases with isotropic scattering and uniform (step) sources. A presentation is first made of the MOC implementation, available in the DRAGON5 code. Two categories of schemes are available for integrating the propagation equations: (1) the first category is based on exact integration and leads to the classical step characteristics (SC) and linear discontinuous characteristics (LDC) schemes and (2) the second category leads to diamond differencing schemes of various orders in space. The acceleration of these MOC schemes using a combination of the generalized minimal residual [GMRES(m)] method preconditioned with the ACA technique was focused on. Numerical results are provided for a two-dimensional (2D) eight-symmetry pressurized water reactor (PWR) assembly mockup in the context of the DRAGON5 code.