• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2343-2351
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• 2024

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.

• Structural Engineering and Mechanics
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• 제2권2호
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• pp.157-171
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• 1994

A general approach to the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced form earthquake. The dynamic responses such as fuel assembly deflected shapes and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.134-140
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• 2017

A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

• Journal of Radiation Protection and Research
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• 제30권1호
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• pp.39-44
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• 2005

노심 및 원자로의 구조 및 구성 물질이 확정되어 있지 않은 개발단계의 신형원자로의 압력용기에 대한 $R-{\theta}$좌표에서 차폐해석을 수행하려면, 매번 선원항에 대한 모델작업을 하는데 많은 노력과 시간이 소요된다. 따라서 $R-\theta$좌표에 의한 반경방향의 원자로 압력용기에 대한 차폐해석에 있어서 노심의 기하학적 구조에 영향을 받지 않고 해석할 수 있는 체계를 개발하였다. 개발된 해석체계를 이용하여 육방형 노심배열을 갖는 일체형 원자로의 압력용기에 대한 차폐해석을 수행하여, 그 결과를 MCNP 해석결과와 비교 분석하였다. 분석결과 개발된 해석체계가 좀 더 보수적인 결과를 나타내었으며 이는 차폐해석측면에서 타당하다. 또한 이 해석체계를 개발함으로써 그 동안 수작업으로 작성하였던 노심내부에 대한 모델에 대한 오차를 줄일 수 있으며 이에 소요되는 시간 및 노력을 줄일 수 있을 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.212-221
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• 1998

In the present study, a method for the dynamic analysis of a reactor core is developed. Peak responses for the motions induced from earthquake are obtained for a core model. The dynamic responses such as fuel assembly shear force, bending moment, axial force and displacement, and spacer grid impact loads are investigated. Prediction of fuel assembly stress during an earthquake requires development of a fuel assembly stress analysis model capable of interfacing with the models and results discussed in the dynamic analysis of a reactor core. This analysis uses beam characteristics which describe the overall fuel assembly response. The stress analysis method and its application for the case of an increased seismic level are also presented.

• 전기의세계
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• 제13권4호
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• pp.16-24
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• 1964

The dynamic equations of the void fraction and the water velocity in boiling region of the BWR reactor core are derived. And these equations are approximated to be able to set on an PACE analog computor. The transient analysis and the frequency response obtained by analog computer are compared with other by digital computor.

• Journal of Radiation Protection and Research
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• 제26권3호
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• Nuclear Engineering and Technology
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• 제50권1호
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.