• 한국통신학회논문지
• /
• 제30권3B호
• /
• pp.138-143
• /
• 2005

본 연구는 CDMA2000 시스템에 있어서의 액세스 프로브 알고리즘을 최적화하고자 한다. 이동국이 액세스 채널을 통하여 송신하고 기지국으로부터 ACK 메시지를 받지 못 할 경우 수신신호가 양호한 지역에서는 PWR_STEP 증분값이 1dEm±0.2 값으로 NUM_STEP만큼 증분하여 액세스 프로브를 하게 된다. 그러나 수신신호가 약한 지역에서는 개방루프 전력제어 알고리즘에 따라 송신부 전력 증폭기가 포화상태에 이르게 되어 PWR_STEP 증분값이 0dBm± 0.2값으로 계속해서 액세스 프로브를 하게 되어 이동국 송신에 따른 간섭현상과 배터리 소비전력이 증가하게 된다. 이와 같은 문제를 해결하기 위하여 우리는 수신신호가 강한 지역에서는 IS-95C규격과 같이 RT 슬롯 값만큼 그리고 수신신호가 미약한 지역에서는 RT+1 슬롯 값만큼 지연하여 액세스 프로브 전력 증분값으로 송신하도록 하여 수신세기에 따라 액세스 프로브 알고리즘을 최적화하였다. 시뮬레이션 결과로부터 제안된 알고리즘이 이동국 송신전력에 따른 간섭현상과 소모 전력을 감소시키고 이로 인해 단말기의 총 통화시간을 향상시킬 수 있음을 확인하였다.

• Nuclear Engineering and Technology
• /
• 제51권2호
• /
• 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
• /
• 제54권6호
• /
• 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
• /
• 제51권5호
• /
• 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.

• 한국정보통신학회논문지
• /
• 제3권4호
• /
• 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
• /
• 제15권2호
• /
• pp.98-109
• /
• 1983

2차원 다군 확산 이론에 의한 원자로 동특성 방정식의 해를 구하기 위해서 two-step alternating direction explicit method를 도입하였다. Alternating direction implicit method의 특별한 경우로써 이 방법의 정확도 및 안전성을 해석하였다. 이 방법의 타당성을 시험하기 위해서 TWIGL 전산조직에 사용한 implicit difference method와 비교하여 두 방법의 결과가 일치함을 알았다. 이 방법을 이용하여 가압경수형 원자로(PWR)의 제어봉 삽입시의 중성자 신속의 시간변화와, CANDU-PHW 원자로의 가상된 냉각재상실 사고시의 중성자 신속의 시간변화를 계산하여 이들 원자로의 제어능력을 확인하였다.

• Nuclear Engineering and Technology
• /
• 제38권1호
• /
• 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.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 추계학술발표회논문집(1)
• /
• 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.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 1993년도 하계학술대회 논문집 A
• /
• 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
• /
• 제49권6호
• /
• 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.