PWR Fuel Management Optimization Using a New Integer Coded Genetic Algorithm

2009 ◽  
Author(s):  
Ahmad Zolfaghari ◽  
Hamid Minuchehr ◽  
Ali Noroozy ◽  
Peymaan Makarachi
Keyword(s):  
Genetic Algorithm ◽  
Objective Function ◽  
Reactor Core ◽  
Complex Problem ◽  
Huge Number ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
Water Reactors ◽  
Management Optimization ◽  
Optimum Solution

The objective of this paper is to develop a new genetic algorithm (GA) for designing the loading pattern (LP) for pressurized water reactors (PWR). Because of huge number of possible combinations for the fuel assemblies (FA’s) loading in a core, finding the optimum solution is truly a complex problem. In common genetic algorithm the mutation and crossover techniques are used to optimize an objective function but in this paper a new modified crossover along a unique technique is presented. In this study flattening of power inside a reactor core is chosen as an objective function. To obtain optimal FA arrangement, a core reload package code, MAKGA, is developed. This code is applicable for all types of PWR core having different geometries and designs with an unlimited number of FA types. The result is well improved in comparison with pattern proposed by designer.

PWR Nuclear Power Plants Fuel Management Optimization

2009 ◽  
Author(s):  
Ahmad Zolfaghari ◽  
Hamid Minuchehr ◽  
Ali Noroozy ◽  
Peymaan Makarachi ◽  
F. Koshahval
Keyword(s):  
Genetic Algorithm ◽  
Objective Function ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Complex Problem ◽  
Pressurized Water ◽  
Water Reactors ◽  
Management Optimization ◽  
Optimum Solution

The objective of this paper is to develop a new hybrid mutation in genetic algorithm (GA) for designing the loading pattern (LP) in pressurized water reactors. Because of huge number of possible combinations for the fuel assemblies (FA’s) loading in a core, finding the optimum solution is truly a complex problem. In common genetic algorithm the mutation and crossover techniques are used to optimize an objective function but in this paper a new hybrid mutation is presented. In this study flattening of power inside a reactor core is chosen as an objective function. To obtain optimal FA arrangement, a core reload package code, MAKNOGA, based on well established MAKGA code is developed. This code is applicable for all types of PWR core having different geometries and designs with an unlimited number of FA types. The result is well improved in comparison with pattern proposed by designer.

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

scholarly journals Optimized Placement of Connecting the Distributed Generationswork Stand Alone to Improve the Distribution Systems Reliability

2013 ◽  
Vol 64 (2) ◽  
pp. 76-83
Author(s):  
Hamed Hashemi-Dezaki ◽  
Ali Agheli ◽  
Behrooz Vahidi ◽  
Hossein Askarian-Abyaneh
Keyword(s):  
Genetic Algorithm ◽  
Objective Function ◽  
System Reliability ◽  
Distribution Systems ◽  
Analytical Approach ◽  
Distributed Generations ◽  
Joint Point ◽  
Cost Of Energy ◽  
Optimum Solution ◽  
The Cost

The use of distributed generations (DGs) in distribution systems has been common in recent years. Some DGs work stand alone and it is possible to improve the system reliability by connecting these DGs to system. The joint point of DGs is an important parameter in the system designing. In this paper, a novel methodology is proposed to find the optimum solution in order to make a proper decision about DGs connection. In the proposed method, a novel objective function is introduced which includes the cost of connector lines between DGs and network and the cost of energy not supplied (CENS) savings. Furthermore, an analytical approach is used to calculate the CENS decrement. To solve the introduced nonlinear optimization programming, the genetic algorithm (GA) is used. The proposed method is applied to a realistic 183-bus system of Tehran Regional Electrical Company (TREC). The results illustrate the effectiveness of the method to improve the system reliability by connecting the DGs work stand alone in proper placements.

AP1000 Plant CILC/CIPS Risk Assessment Using Advanced TH Methodology

2012 ◽  
Author(s):  
G. Wang ◽  
P. Sapienza ◽  
R. J. Fetterman ◽  
M. Y. Young ◽  
J. R. Secker ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Nucleate Boiling ◽  
Operating Conditions ◽  
Localized Corrosion ◽  
Pressurized Water ◽  
Assembly Design ◽  
Power Shift ◽  
Fuel Assemblies ◽  
Water Reactors ◽  
The Relationship

Similar to many existing Pressurized Water Reactors (PWR), the AP1000® cores will undergo sub-cooled nucleate boiling in the upper grid spans of some fuel assemblies at normal operating conditions. Sub-cooled nucleate boiling may increase crud deposits on the fuel cladding surface which may increase the risk of Crud Induced Power Shift (CIPS) and/or Crud Induced Localized Corrosion (CILC). A CIPS/CILC risk assessment has been performed to support the AP1000 fuel assembly design finalization. In this paper, the advanced thermal-hydraulic (TH) methodology used in the AP1000 plant CIPS/CILC risk assessments are summarized and discussed, and the relationship between the CIPS/CILC mechanisms, fuel reliability, and plant operating conditions is also presented. Finally, acceptable AP1000 core CIPS/CILC risk assessment results are summarized and suggestions that specifically target reducing CIPS/CILC risks for AP1000 plants are described.

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Analysis of Thermal-Hydraulic Fluctuations in Trillo NPP With CTF/PARCS v. 2.7 Coupled Code

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
A. Abarca ◽  
R. Miró ◽  
G. Verdú ◽  
J. A. Bermejo
Keyword(s):  
Reactor Core ◽  
Low Frequency ◽  
Protection System ◽  
Inlet Temperature ◽  
Frequency Noise ◽  
Pressurized Water ◽  
The Core ◽  
Coupled Codes ◽  
Water Reactors ◽  
New Research

The low-frequency noises are fluctuations in the neutron flux density, in the low-frequency range up to 4 Hz, which generate noise in the neutron instrumentation and could affect the limitation and protection system of the reactor core. Some European pressurized water reactors (PWRs) experienced the effect of low-frequency noise, opening a new research line for the verification of the neutron-kinetics/thermal-hydraulic coupled codes. A CTF/PARCS v. 2.7 simulation study to verify whether periodical fluctuations in the core inlet temperature could activate the core protection system has been done, obtaining the frequency spectrum of the power oscillation amplitudes.

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