Neural network and genetic algorithms for optimizing the plate element of Egyptian research reactor problems

The Second Egyptian Research Reactor ET-RR-2 went critical on the 27th of November 1997. The National Center of Nuclear Safety and radiation Control ‘NCNSRC’ has the responsibility of the evaluation and the assessment of the safety of this reactor. Multi-objective optimization is a powerful tool for resolving conflicting objectives in engineering design and numerous other fields. The purpose of this paper is to present an approach to the optimization of the fuel element plate, which is designed with a view to improve reliability and lifetime and it is one of the most important elements during the shut down. In this paper, we present a conceptual design approach for fuel element plate, in conjunction with a genetic algorithm and comparing with neural networks to obtain a fuel plate that maximizes a fitness value to optimize the safety design of the fuel plate.     

Parameter optimization in molecular dynamics simulations using a genetic algorithm

In this work, we introduce a genetic algorithm for the parameterization of the reactive force field developed by Kieffer [12], [13], [14], [15] and [16]. This potential includes directional covalent bonds and dispersion terms. Important features of this force field for simulating systems that undergo significant structural reorganization are (i) the ability to account for the redistribution of electron density upon ionization, formation, or breaking of bonds, through a charge transfer term, and (ii) the fact that the angular constraints dynamically adjust when a change in the coordination number of an atom occurs.In this paper, we present the implementation of the genetic algorithm into the existing code as well as the algorithm efficiency and preliminary results on Si-Si force field optimization. The parameters obtained by this method will be compared to existing parameter sets obtained by a trial-and-error process.     

基于遗传算法的自由电子激光优化设计

自由电子激光(Free-Electron Laser,FEL)的辐射功率、光谱等关键量是表征FEL品质的重要因素,这些量往往依赖于多种参量,所以优化这些品质参量的问题即可等效为如何寻求合适的参数来获得更优的FEL的输出。遗传算法是解决这类多变量优化问题常用的算法之一。本文基于遗传算法设计了一个用于FEL优化的应用程序,该应用程序利用实数编码方式,选择合适的算子并作相应的改进,同时利用Java Swing构建了友好的用户界面。实验结果表明,在进行辐射功率优化时,该算法能够在较短的时间内寻找到非常接近全局最优解的较优解。该应用程序具有良好的通用性与可扩展性,在一定程度上为FEL装置的运行优化提供帮助。     

Seismic response analysis of full-scale boiling water reactor using three-dimensional finite element method

This paper presents the three-dimensional finite element seismic response analysis of full-scale boiling water reactor BWR5 at Kashiwazaki-Kariwa Nuclear Power Station subjected to the Niigata-ken Chuetsu-Oki earthquake that occurred on 16 July 2007. During the earthquake, the automatic shutdown system of the reactors was activated successfully. Although the monitored seismic acceleration significantly exceeded the design level, it was found that there were no significant damages of the reactor cores or other important systems, structures and components through in-depth investigation. In the seismic design commonly used in Japan, a lumped mass model is employed to evaluate the seismic response of structures and components. Although the lumped mass model has worked well so far for a seismic proof design, it is still needed to develop more precise methods for the visual understanding of response behaviors. In the present study, we propose the three-dimensional finite element seismic response analysis of the full-scale and precise BWR model in order to directly visualize its dynamic behaviors. Through the comparison between both analysis results, we discuss the characteristics of both models. The stress values were also found to be generally under the design value.     

Design of an intelligent fuzzy logic controller for a nuclear research reactor

The main objective of this paper is to design an intelligent controller system based on the concepts of fuzzy logic. This latter will be used to control the power of a nuclear reactor. The principle of this controller is based on rules established from experiments used with a classical controller and from the knowledge and the expertise of the operators of the reactor. This intelligent controller could be used in parallel with the actual system, which is semiautomatic, as a decision aided system to assist the operators in the control room.     

广义预测自校正控制算法在核反应堆功率控制中的应用研究

阐述了核反应堆堆芯的中子动力学数学模型,并应用该模型建立传递函数.提出将广义预测自校正控制算法应用于反应堆功率控制中,包括控制结构和控制器设计.仿真结果表明所采用的广义预测自校正控制算法能够较好地控制反应堆功率的输出,取得了较好的控制效果.     

遗传算法在高通量工程试验堆燃料管理优化中的应用

建立了基于遗传算法的高通量工程试验堆(HFETR)堆内燃料管理优化模型。根据HFETR的实际情况，提出了基于组件位置的二进制编码/解码技术。研究了不同选择策略对算法性能的影响，探讨了遗传算法和专家经验的结合，吸收了自适应遗传算法的思想，得到了可直接应用于HFETR的优化换料方案。     

A class-based search for the in-core fuel management optimization of a pressurized water reactor

The In-Core Fuel Management Optimization (ICFMO) is a prominent problem in nuclear engineering, with high complexity and studied for more than 40 years. Besides manual optimization and knowledge-based methods, optimization metaheuristics such as Genetic Algorithms, Ant Colony Optimization and Particle Swarm Optimization have yielded outstanding results for the ICFMO. In the present article, the Class-Based Search (CBS) is presented for application to the ICFMO. It is a novel metaheuristic approach that performs the search based on the main nuclear characteristics of the fuel assemblies, such as reactivity. The CBS is then compared to the one of the state-of-art algorithms applied to the ICFMO, the Particle Swarm Optimization. Experiments were performed for the optimization of Angra 1 Nuclear Power Plant, located at the Southeast of Brazil. The CBS presented noticeable performance, providing Loading Patterns that yield a higher average of Effective Full Power Days in the simulation of Angra 1 NPP operation, according to our methodology.     

高温气冷实验堆燃料元件双向探测器的研制

介绍了高温气冷实验堆燃料元件双向探测器的基本原理和实现方法。它以两个并联的感应线圈为敏感元件,通过双通道法采集信号,以８９Ｃ５１单片机为处理核心,系统软件采用循环扫描输入端口的方式获取过球信号,经智能分析、判断,实现了燃料元件的双向检测。     

反应堆压力壳(RV)螺栓孔检查设备的研制

介绍了一种新型的能在高辐射环境下工作的全自动检查反应堆压力壳法兰主螺栓孔螺纹缺陷的设备一螺纹检查机器人。对检查设备的性能参数、技术难点进行了分析。     

A thermal model to characterize the flattening effect of a nuclear fuel element in an annular channel using simple analytical approach

In the present work, axial power flattening effect due to insertion of adjuster rods is modeled and a heat generating fuel pin in an annular channel is characterized. The effect of insertion of single adjustor rod at the peak heat generation region is modeled as two heat sources and one heat sink. The results obtained from the developed model indicate that insertion of adjuster rods beyond a critical number (3–4); flattening effect does not improve much. The peak surface and centerline temperatures of fuel pin are observed to show a cyclic variation and shift towards end as compared to unadjusted profile. Results are in accordance with the best estimate RELAP code.     

Study on the characteristic points of boiling curve by using wavelet analysis and genetic neural network

Local singularity of a signal includes a lot of important information. Wavelet transform can overcome the shortages of Fourier analysis, i.e., the weak localization in the local time- and frequency-domains. It has the capacity to detect the characteristic points of boiling curves. Based on the wavelet analysis theory of signal singularity detection, Critical Heat Flux (CHF) and Minimum Film Boiling Starting Point (q_min) of boiling curves can be detected by using the wavelet modulus maxima detection. Moreover, a genetic neural network (GNN) model for predicting CHF is set up in this paper. The database used in the analysis is from the 1960s, including 2365 data points which cover a range of pressure (P), from 100 to 1000 kPa, mass flow rate (G) from 40 to 500 kg m⁻² s⁻¹, inlet sub-cooling (ΔT_sub) from 0 to 35 K, wall superheat (ΔT_sat) from 10 to 500 K and heat flux (Q) from 20 to 8000 kW m⁻². GNN mode has some advantages of its global optimal searching, quick convergence speed and solving non-linear problem. The methods of establishing the model and training of GNN are discussed particularly. The characteristic point predictions of boiling curve are investigated in detail by GNN. The results predicted by GNN have a good agreement with experimental data. At last, the main parametric trends of the CHF are analyzed by applying GNN. Simulation and analysis results show that the network model can effectively predict CHF.     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Seismic analysis of the APR 1400 reactor vessel internals using the model reduction method

To secure reliability of the seismic design of the reactor vessel internals (RVIs) through the finite element analysis, it is important to develop the accurate analysis model which can represent the geometric complexity of the RVIs. However, the seismic analysis requires too large computation cost to solve the complex equations; thus, it needs to reduce the overall size of the analysis model. Here, we apply a model reduction method based on the fixed-interface component mode synthesis (CMS) method to practical RVIs to solve complex numerical problems efficiently. To verify the model reduction method, several cases of the RVIs with different conditions are analyzed for the static and dynamic problems. Finally, the seismic analysis was performed with the suggested reduced model with the CMS method. The time history analysis is performed to extract important seismic responses at the specified locations, and the stress analysis is also performed to identify that the RVIs satisfy the seismic design. In the last part of the paper, an example of the design modification is suggested to reduce the stress intensity at the support locations.     

Developing a practical optimization of the refueling program for ordinary research reactors using a modified simulated annealing method

Each core configuration of a research reactor can be optimized to provide particular or general multi-purpose irradiating conditions; specially, it includes refueling cycle length and irradiating neutron fluxes if a core management is limited to a refueling task. Each practical core or fuel management operation needs providing all of related Operational Limits and safety Conditions (OLCs). In this paper refueling cycle length and maximum irradiating thermal neutron flux are chosen as the optimizing objectives; also OLCs including total Power Peaking Factor, Shutdown Margin, Reactivity Safety Factor (RSF), and maximum permissible core excess reactivity are influenced as optimizing constraints. All parameters have been calculated accurately and benchmarked against operational parameters of a 5 MW MTR. Primary 2-D annealing process is following up to a secondary re-annealing in fine 3-D calculations. This expands global search space while the required time is reduced. Safety margins are introduced by stepwise penalty functions instead of a direct rejecting method. Results are very promising, required iterations are decreased; safety faults are automatically removed, and final results are gained near touch the infeasible frontier formed by safety margins. Refueling cycle length is significantly increased, averaged and maximum irradiating neutron fluxes are enhanced while selected OLCs are passed.     

A study of the decommissioning procedure of an activated structure through an evaluation of the decommissioning cost for a research reactor

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan.This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor.As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research.     

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents.Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.     

New research programme of JAEA/CLADS to reduce the knowledge gaps revealed after an accident at Fukushima-1: introduction of boiling water reactor mock-up assembly degradation test programme

ABSTRACT

The new R&D programme of JAEA/CLADS tests complements the previous investigations related to BWR severe accidents. A series of tests aim at closing the gaps in understanding of the Fukushima Dai-Ichi degradation sequence at each unit. The paper emphasises the problem of control blade degradation, which influences the accident progression at an early stage and shows the approach for thorough investigation of this problem.     

An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR

To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.