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.     

Multiobjective fuel management optimization for self-fuel-providing LMFBR using genetic algorithms

One of the conceptual options under consideration for the future of nuclear power is the long-term development without fuel reprocessing. This concept is based on a reactor that requires no plutonium reprocessing for itself, and provides high efficiency of natural uranium utilization, so called Self-Fuel-Providing LMFBR (SFPR). Several design considerations were previously given to this reactor type which, however, suffer from some problems connected with insufficient power flattening, large reactivity swings during burnup cycles, and peak fuel burnup being significantly higher than recent technology experience, which is about 18% for U-10 wt%Zr metallic fuel to be considered. Yet, the mentioned core parameters demonstrate high sensitivity to the fuel management strategy selected for the reactor. Therefore, the aim of this study is to develop a practical tool for the improvement of the core characteristics by fuel management optimization, which is based on advanced optimization techniques such as Genetic Algorithms (GA). The calculation results obtained by a simplified reactor model can serve as estimates of achievable values for mentioned core parameters, which are necessary to make decisions at the preliminary optimization stage.     

Application of genetic algorithms to optimize burnable poison placement in pressurized water reactors

An efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poison (BP) placement optimization problem in the reference Three Mile Island-1 (TMI-1) core. Core BP optimization problem means developing a BP loading map for a given core loading pattern that minimizes the total Gadolinium (Gd) amount in the core without violating any design constraints. The number of UO₂/Gd₂O₃ pins and Gd₂O₃ concentrations for each fresh fuel location in the core are the decision variables. The objective function was to minimize the total amount of Gd in the core together with the residual Gd reactivity binding at the End-of-Cycle (EOC). The constraints are to keep the maximum peak pin power during the core depletion and soluble boron (SOB) concentration at the Beginning of Cycle (BOC) both less than their limit values. The innovation of this study was to search all of the possible UO₂/Gd₂O₃ fuel assembly designs with variable number of UO₂/Gd₂O₃ fuel pins and concentration of Gd₂O₃ in the overall decision space. The use of different fitness functions guided the solution towards desired (good solutions) region in the solution space, which accelerated the GA solution. The main objective of this study was to develop a practical and efficient GA tool and to apply this tool to designing an optimum BP pattern for a given core loading.     

AXIAL: a system for boiling water reactor fuel assembly axial optimization using genetic algorithms

A system named AXIAL is developed based on the genetic algorithms (GA) optimization method, using the 3D steady state simulator code Core-Master-PRESTO (CM-PRESTO) to evaluate the objective function. The feasibility of this methodology is investigated for a typical boiling water reactor (BWR) fuel assembly (FA). The axial location of different fuel compositions is found in order to minimize the FA mean enrichment needed to obtain the cycle length under the safety constraints. Thermal limits are evaluated at the end of cycle using the Haling calculation; the hot excess reactivity and the shutdown margin at the beginning of cycle are also evaluated. The implemented objective function is very flexible and complete, incorporating all the thermal and reactivity limits imposed during fuel design analysis; furthermore, additional constraints can be easily introduced in order to obtain an improved solution. The results show a small improvement in the FA average enrichment obtained with the system related to the reference case that has been studied. The results show that the system converge to an optimal solution, it is observed that the mean fuel enrichment decreases while all the constraints are satisfied. A comparison was also performed using one-point and two-points crossover operator and the results of a sensitivity study for different mutation percentage are also showed.     

Research reactor in-core fuel management optimization by application of multiple cyclic interchange algorithms

Fuel shuffling optimization procedures are proposed for the Hoger Onderwijs Reactor (HOR) in Delft, The Netherlands, a 2MWth swimming-pool type research reactor. These procedures are based on the multiple cyclic interchange approach, according to which the search for the reload pattern associated with the highest objective function value can be thought of as divided in multiple stages. The transition from the initial to the final stage is characterized by an increase in the degree of locality of the search procedure. The general idea is that, during the first stages, the ‘elite’ cluster containing the group of best patterns must be located, after which the solution space is sampled in a more and more local sense to find the local optimum in this cluster. The transition(s) from global search behaviour to local search behaviour can be either prompt, by defining strictly separate search regimes, or gradual by introducing stochastic acceptance tests. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

A non-equilibrium cycle approach for optimization of successive nuclear fuel reloads in PWR's

A non-equilibrium cycle approach for PWR fuel management has been developed based on fuel cycle and plant operation experience. This approach is incorporated into a computer code. MISER. The methods employed give emphasis to the accuracy of physics modeling and realistic reactor operating constraints. The non-equilibrium approach provides a flexible and practical method of evaluating cost-optimal short- and long-term fuel policies in the current situation of changing fuel economics and reactor operating conditions. In particular, it is being applied to fuel cycle optimization for San Onofre Unit 1, and, in general has the capability of performing fuel management on larger commercial PWR's (see Lew, 1976; Lew and Fenech, 1976).For San Onofre Unit 1, a non-equilibrium policy was found to have a potential 4% lower levelized fuel cycle costs and a potential 15% lower initial capital costs than the current equilibrium fuel management policy. This represents a potential fuel cycle cost saving on the order of nearly a half million dollars and potential initial capital cost savings of a million dollars per yr. This minimum cost policy was found to be sensitive to the assumptions of fuel salvage value. For the policies and operating constraints considered, the minimum costs policy was found to be insensitive to the range of U₃O₈ and enrichment costs considered. In addition, this policy was also found to be insensitive to variations in the adjustable physics power and burn-up parameters and to reductions in the maximum radial peaking factor constraint.     

A systematic approach to the radiation damage problem

To assess the suitability of a material for use as a core component in a fast reactor or for the first wall in a fusion reactor, it is necessary to know the irradiation damage behaviour of the material outside the usual materials testing data domain. In the present paper we propose a strategy based on a closely co-ordinated programme of experimental and theoretical research. The aim of this strategy is the systematic construction of a physically based model of the evolving damage structures. This would then allow both the necessary extrapolations of the data to the desired conditions to be achieved in a reliable fashion and provide a rational basis for the development of low-swelling alloys for the two nuclear systems.     

Ensuring the possibility of using thorium as a fuel in a pressurized water reactor(PWR)

The possibility of utilizing thorium as a fuel in a pressurized water reactor(PWR)has been proven from the neutronic perspective in our previously published work without assessing the thermal hydraulic(TH)and solid structure performances.Therefore,the TH and solid structure performances must be studied to confirm these results and ensure the possibility of using a thorium-based fuel as an excellent accident-tolerant fuel.The TH and solid structure performances of thorium-based fuels were investigated and compared with those of U02.The radial and axial power peaking factors(PPFs)for U02,(232Th,235U)02,and(232Th,233U)02 were examined with a PWR assembly to determine the total PPF of each one.Both Gd203 and Er203 were tested as burnable absorbers(BAs)to manage the excess reactivity at the beginning of the fuel cycle(BOC)and reduce the total PPF.Er203 resulted in a more significant reduction to the total PPF and,therefore,a greater reduction to the temperature distribution compared to Gd203.Given these results,we analyzed the effects of adding Er203 to thorium-based fuels on their TH and solid structure performances.     

Application of large-eddy simulation to pressurized thermal shock problem: A grid resolution study

Life time extension assessment is a very important issue for the nuclear community. A serious threat to the life time extension of a Reactor Pressure Vessel (RPV) is an occurrence of the Pressurized Thermal Shock (PTS) during an Emergency Core Coolant (ECC) injection in a loss-of-coolant accident. Traditional system codes fail to predict the complex three-dimensional flow phenomena resulting from such ECC injection. Improved results have been obtained using Computational Fluid Dynamics (CFD) analysis based on the Reynolds-Averaged Navier-Stokes (RANS) equations. However, it has been also shown that transient RANS approaches are less capable to predict the complex flow features in the downcomer of the RPV. More advanced CFD methods like Large-Eddy Simulation (LES) are required for modeling of these flow phenomena. This paper addresses the feasibility of the application of LES for single-phase PTS. Furthermore, the required grid resolution for such LES analyses is identified by evaluation of the solution on different mesh sizes. A buoyancy-driven PTS experiment has been considered here. This experiment has been performed in the 1:5 linear scale Rossendorf Coolant Mixing Model (ROCOM) facility. In the applied numerical model, the incompressible Navier-Stokes equations are solved in the LES formulation, with an additional transport equation for a scalar, which is responsible for driving the embedded buoyancy term in the momentum equations. Instantaneous mixing characteristics are investigated based on evaluation of the scalar concentration. The analysis presented in this paper indicates that the application of LES is feasible nowadays for single-phase PTS. It is demonstrated that the mixing in the downcomer is quite sensitive to small turbulent disturbances at the ECC inlet, i.e., two simulations performed with slightly different fluctuations at the inlet result in substantially different flow in the downcomer. This complicates the analysis of the data from simulations and suggests that evaluation of the results should be performed in the frequency-amplitude domain instead of the classically employed temporal data analysis.     

基于遗传算法的压水堆核电一回路稳压器机理建模与仿真

针对压水堆核电站一回路稳压器实际运行特性,根据能量守恒、质量守恒和动量守恒方程,考虑了喷淋流量、电加热器功率及安全释放阀的影响,建立了一个两相动态非平衡的稳压器机理模型.为提高模型精度,采用遗传算法对该模型的参数进行优化,用以得到一组模型的最优参数.将参数优化算法应用于某900 MW核电站稳压器仿真实例,通过模块封装组建成稳压器压力仿真模型,并与核电厂提供的对应数据做了比较验证了建模方法的正确性及优化方法的有效性.     

A two-dimensional thermohydraulic calculation on the core of pressurized water reactor

Nuclear Research and Nuclear Power Institute, Bulgarian Academy of Sciences. Translated from Atomnaya Énergiya, Vol. 73, No. 5, pp. 397–400, November, 1992.     

Optimization of the power ascension path for a boiling water reactor using genetic algorithms

To ensure the core status can meet the requirements of thermal limits, stability and other constraints during the power ascension process of a nuclear power plant, operators usually gradually increase power based on onsite measurements and experience. To reduce the operator’s burden, this research develops a method to find an optimal power ascension path that can be followed by operators. The power ascension path is formulated as a multiobjective optimization problem with the following constraints: power ascension time, thermal limits, core stability and maximum rod line. A genetic algorithm is adopted to obtain the optimal power ascension path. The results show that using our approach full power can be achieved quickly, while maintaining reasonable margins of thermal limit and stability, in addition to satisfying maximum rod line criteria.     

Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

1 Introduction With respect to the inherent safety of nuclear re- actors, application of passive systems/components including natural circulation phenomena as the main mechanism is intended to simplify the safety-related systems and to improve their reliability, to reduce the effect of human errors and equipment failures, and to provide more time to enable the operators to prevent or mitigate serious accidents. Natural circulation is the main mode of heat removal for removing decay heat from t…     

A risk approach to the management of boiler tube thinning

A large set of industrial thickness inspection data covering four boiler units of a power station over a period of five years was made available to the authors. The measurements were made in regions of the boiler where corrosion/erosion was the major cause of failure of the boiler tubes. There were over 40,000 separately measured data points in the data and all were collected with some care and expense.In the development of maintenance strategies for equipment, this type of data is typical of the data that must be collected and assessed. This data thus represents an opportunity to evaluate the ability to generate a useful risk approach to the management of the tubing.An important example of a risk-based approach is the American Petroleum Institute (API) Risk Based Inspection (“RBI”), API 581. A variety of problems were encountered applying this to boiler tubes. The problems include irrelevant API 581 corrosion rate tables, lack of information on how to analyse inspection data, difficulty of dealing with multiple inspection categories and lack of suitable direction for programming inspection intervals.     

Solving the inverse problem of parameter estimation by genetic algorithms: the case of a groundwater contaminant transport model

Mathematical models are widely used within the performance assessment of radioactive waste repositories to describe the behaviour of groundwater systems under the various physical conditions encountered throughout the long time scales involved. The effectiveness of such predictive models largely depends on the accuracy with which the involved parameters can be determined. In the present paper, we investigate the feasibility of using genetic algorithms for estimating the parameters of a groundwater contaminant transport model. The genetic algorithms are numerical search tools aiming at finding the global optimum of a given real objective function of one or more real variables, possibly subject to various linear or non linear constraints. The search procedures provided by the genetic algorithms resemble certain principles of natural evolution. In the case study here, the transport of contaminants through a three-layered monodimensional saturated medium is numerically simulated by a monodimensional advection–dispersion model. The associated velocity and dispersivity parameters are estimated by a genetic algorithm whose objective function is the sum of the squared residuals between pseudo-experimental data, obtained with the true values of the parameters, and the concentration profiles computed with the model using the estimated values of the parameters. The results indicate that the method is capable of estimating the parameters values with accuracy, also when in presence of substantial noise. Furthermore, we investigate the possibility of extracting some qualitative information regarding the sensitivity of the model to the unknown input parameters from the speed of convergence and stabilization of the identification procedure.     

Active measures to preclude breaks in the reactor coolant lines of pressurized water reactors

The active measures which preclude a spontaneous failure of the reactor coolant lines are exemplarily shown. With the basic safety concept a quality standard is achieved characterized by high-grade material properties, a structure that is adequate to the loads to which the components will be subjected in service and is amenable to inspection, precise load and stress evaluation, optimized manufacturing and operation monitoring. The possible failure types are described and the safety against failure is assessed.