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Sample records for nuclear core optimization

  1. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto

    2005-01-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  2. Improving the calculated core stability by the core nuclear design optimization

    International Nuclear Information System (INIS)

    Partanen, P.

    1995-01-01

    Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)

  3. In-core nuclear fuel management optimization of VVER1000 using perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2011-01-01

    In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

  4. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Schirru, Roberto; Carvalho da Silva, Fernando; Medeiros, Jose Antonio Carlos Canedo

    2008-01-01

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem

  5. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  6. A system for obtaining an optimized pre design of nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1989-01-01

    This work proposes a method for obtaing a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one-energy-group, unidimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, refletor thickness, enrichement and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for futures works. (author) [pt

  7. A system to obtain an optimized first design of a nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, L.A.

    1988-01-01

    This work proposes a method for obtaining a first design of nuclear reactor cores. It takes into consideration the objectives of the project, physical limits, economical limits and the reactor safety. For this purpose, some simplifications were made in the reactor model: one energy-group, one-dimensional and homogeneous core. The adopted model represents a typical PWR core and the optimized parameters are the fuel thickness, reflector thickness, enrichment and moderating ratio. The objective is to gain a larger residual reactivity at the end of the cycle. This work also presents results for a PWR core. From the results, many conclusions are established: system efficiency, limitations and problems. Also some suggestions are proposed to improve the system performance for future works. (autor)

  8. Evaluation of the need for stochastic optimization of out-of-core nuclear fuel management decisions

    International Nuclear Information System (INIS)

    Thomas, R.L. Jr.

    1989-01-01

    Work has been completed on utilizing mathematical optimization techniques to optimize out-of-core nuclear fuel management decisions. The objective of such optimization is to minimize the levelized fuel cycle cost over some planning horizon. Typical decision variables include feed enrichments and number of assemblies, burnable poison requirements, and burned fuel to reinsert for every cycle in the planning horizon. Engineering constraints imposed consist of such items as discharge burnup limits, maximum enrichment limit, and target cycle energy productions. Earlier the authors reported on the development of the OCEON code, which employs the integer Monte Carlo Programming method as the mathematical optimization method. The discharge burnpups, and feed enrichment and burnable poison requirements are evaluated, initially employing a linear reactivity core physics model and refined using a coarse mesh nodal model. The economic evaluation is completed using a modification of the CINCAS methodology. Interest now is to assess the need for stochastic optimization, which will account for cost components and cycle energy production uncertainties. The implication of the present studies is that stochastic optimization in regard to cost component uncertainties need not be completed since deterministic optimization will identify nearly the same family of near-optimum cycling schemes

  9. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura, E-mail: ameneses@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program; Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Gambardella, Luca Maria, E-mail: luca@idsia.c [Dalle Molle Institute for Artificial Intelligence (IDSIA), Manno-Lugano, TI (Switzerland); Schirru, Roberto, E-mail: schirru@lmp.ufrj.b [COPPE/Federal University of Rio de Janeiro, RJ (Brazil). Nuclear Engineering Program

    2009-07-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  10. A nuclear heuristic for application to metaheuristics in-core fuel management optimization

    International Nuclear Information System (INIS)

    Meneses, Anderson Alvarenga de Moura; Gambardella, Luca Maria; Schirru, Roberto

    2009-01-01

    The In-Core Fuel Management Optimization (ICFMO) is a well-known problem of nuclear engineering whose features are complexity, high number of feasible solutions, and a complex evaluation process with high computational cost, thus it is prohibitive to have a great number of evaluations during an optimization process. Heuristics are criteria or principles for deciding which among several alternative courses of action are more effective with respect to some goal. In this paper, we propose a new approach for the use of relational heuristics for the search in the ICFMO. The Heuristic is based on the reactivity of the fuel assemblies and their position into the reactor core. It was applied to random search, resulting in less computational effort concerning the number of evaluations of loading patterns during the search. The experiments demonstrate that it is possible to achieve results comparable to results in the literature, for future application to metaheuristics in the ICFMO. (author)

  11. The Great Deluge Algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Oliveira, Cassiano R.E. de

    2005-01-01

    The Great Deluge Algorithm (GDA) is a local search algorithm introduced by Dueck. It is an analogy with a flood: the 'water level' rises continuously and the proposed solution must lie above the 'surface' in order to survive. The crucial parameter is the 'rain speed', which controls convergence of the algorithm similarly to Simulated Annealing's annealing schedule. This algorithm is applied to the reactor core design optimization problem, which consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. This problem was previously attacked by the canonical genetic algorithm (GA) and by a Niching Genetic Algorithm (NGA). NGAs were designed to force the genetic algorithm to maintain a heterogeneous population throughout the evolutionary process, avoiding the phenomenon known as genetic drift, where all the individuals converge to a single solution. The results obtained by the Great Deluge Algorithm are compared to those obtained by both algorithms mentioned above. The three algorithms are submitted to the same computational effort and GDA reaches the best results, showing its potential for other applications in the nuclear engineering field as, for instance, the nuclear core reload optimization problem. One of the great advantages of this algorithm over the GA is that it does not require special operators for discrete optimization. (author)

  12. Optimization analysis of the nuclear fuel cycle transition to the last core

    International Nuclear Information System (INIS)

    Rebollo, L.; Blanco, J.

    2001-01-01

    The Zorita NPP was the first Spanish commercial nuclear reactor connected to the grid. It is a 160 MW one loop PWR, Westinghouse design, owned by UFG, in operation since 1968. The configuration of the reactor core is based on 69 fuel elements type 14 x 14, the standard reload of the present equilibrium cycle being based on 16 fuel elements with 3.6% enrichment in 235 U. In order to properly plan the nuclear fuel management of the transition cycles to its end of life, presently foreseen by 2008, an based on the non-reprocessing option required by the policy of the Spanish Administration, a technical-economical optimization analysis has been performed. As a result, a fuel management strategy has been defined looking for getting simultaneously the minimum integral fuel cost of the transition from the present equilibrium cycle to the last core, as well as the minimum residual worth of the fuel remaining in the core after the final outage. Based on the ''lessons learned'' derived from the study, the time margin for the decision making has been determined, and a planning of the nuclear fuel supply for the transition reloads, specifying both the number of fuel elements and their enrichment in 235 U, as been prepared. Finally, based on the calculated economical worth of the partially burned fuel of the last core, after the end of its operation cycle, a financial cover for yearly compensation from now on of the foreseen final lost has been elaborated. Most of the conceptual conclusions obtained are applicable to the other commercial nuclear reactors in operation owned by UFG, so that they are understood to be of general interest and broad application to commercial PWR. (author)

  13. Development of a parallel genetic algorithm using MPI and its application in a nuclear reactor core. Design optimization

    International Nuclear Information System (INIS)

    Waintraub, Marcel; Pereira, Claudio M.N.A.; Baptista, Rafael P.

    2005-01-01

    This work presents the development of a distributed parallel genetic algorithm applied to a nuclear reactor core design optimization. In the implementation of the parallelism, a 'Message Passing Interface' (MPI) library, standard for parallel computation in distributed memory platforms, has been used. Another important characteristic of MPI is its portability for various architectures. The main objectives of this paper are: validation of the results obtained by the application of this algorithm in a nuclear reactor core optimization problem, through comparisons with previous results presented by Pereira et al.; and performance test of the Brazilian Nuclear Engineering Institute (IEN) cluster in reactors physics optimization problems. The experiments demonstrated that the developed parallel genetic algorithm using the MPI library presented significant gains in the obtained results and an accentuated reduction of the processing time. Such results ratify the use of the parallel genetic algorithms for the solution of nuclear reactor core optimization problems. (author)

  14. The gravitational attraction algorithm: a new metaheuristic applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Oliveira, Cassiano R.E. de

    2005-01-01

    A new metaheuristic called 'Gravitational Attraction Algorithm' (GAA) is introduced in this article. It is an analogy with the gravitational force field, where a body attracts another proportionally to both masses and inversely to their distances. The GAA is a populational algorithm where, first of all, the solutions are clustered using the Fuzzy Clustering Means (FCM) algorithm. Following that, the gravitational forces of the individuals in relation to each cluster are evaluated and this individual or solution is displaced to the cluster with the greatest attractive force. Once it is inside this cluster, the solution receives small stochastic variations, performing a local exploration. Then the solutions are crossed over and the process starts all over again. The parameters required by the GAA are the 'diversity factor', which is used to create a random diversity in a fashion similar to genetic algorithm's mutation, and the number of clusters for the FCM. GAA is applied to the reactor core design optimization problem which consists in adjusting several reactor cell parameters in order to minimize the average peak-factor in a 3-enrichment-zone reactor, considering operational restrictions. This problem was previously attacked using the canonical genetic algorithm (GA) and a Niching Genetic Algorithm (NGA). The new metaheuristic is then compared to those two algorithms. The three algorithms are submitted to the same computational effort and GAA reaches the best results, showing its potential for other applications in the nuclear engineering field as, for instance, the nuclear core reload optimization problem. (author)

  15. Development of a multi-objective PBIL evolutionary algorithm applied to a nuclear reactor core reload optimization problem

    International Nuclear Information System (INIS)

    Machado, Marcelo D.; Dchirru, Roberto

    2005-01-01

    The nuclear reactor core reload optimization problem consists in finding a pattern of partially burned-up and fresh fuels that optimizes the plant's next operation cycle. This optimization problem has been traditionally solved using an expert's knowledge, but recently artificial intelligence techniques have also been applied successfully. The artificial intelligence optimization techniques generally have a single objective. However, most real-world engineering problems, including nuclear core reload optimization, have more than one objective (multi-objective) and these objectives are usually conflicting. The aim of this work is to develop a tool to solve multi-objective problems based on the Population-Based Incremental Learning (PBIL) algorithm. The new tool is applied to solve the Angra 1 PWR core reload optimization problem with the purpose of creating a Pareto surface, so that a pattern selected from this surface can be applied for the plant's next operation cycle. (author)

  16. A hybrid niched-island genetic algorithm applied to a nuclear core optimization problem

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.

    2005-01-01

    Diversity maintenance is a key-feature in most genetic-based optimization processes. The quest for such characteristic, has been motivating improvements in the original genetic algorithm (GA). The use of multiple populations (called islands) has demonstrating to increase diversity, delaying the genetic drift. Island Genetic Algorithms (IGA) lead to better results, however, the drift is only delayed, but not avoided. An important advantage of this approach is the simplicity and efficiency for parallel processing. Diversity can also be improved by the use of niching techniques. Niched Genetic Algorithms (NGA) are able to avoid the genetic drift, by containing evolution in niches of a single-population GA, however computational cost is increased. In this work it is investigated the use of a hybrid Niched-Island Genetic Algorithm (NIGA) in a nuclear core optimization problem found in literature. Computational experiments demonstrate that it is possible to take advantage of both, performance enhancement due to the parallelism and drift avoidance due to the use of niches. Comparative results shown that the proposed NIGA demonstrated to be more efficient and robust than an IGA and a NGA for solving the proposed optimization problem. (author)

  17. A nuclear reactor core fuel reload optimization using Artificial-Ant-Colony Connective Networks; Recarga de reatores nucleares utilizando redes conectivas de colonias de formigas artificiais

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: alan@lmp.ufrj.br; schirru@lmp.ufrj.br

    2005-07-01

    A Pressurized Water Reactor core must be reloaded every time the fuel burnup reaches a level when it is not possible to sustain nominal power operation. The nuclear core fuel reload optimization consists in finding a burned-up and fresh-fuel-assembly pattern that maximizes the number of full operational days. This problem is NP-hard, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Besides that, the problem is non-linear and its search space is highly discontinual and multimodal. In this work a parallel computational system based on Ant Colony System (ACS) called Artificial-Ant-Colony Networks is introduced to solve the nuclear reactor core fuel reload optimization problem. ACS is a system based on artificial agents that uses the reinforcement learning technique and was originally developed to solve the Traveling Salesman Problem, which is conceptually similar to the nuclear fuel reload problem. (author)

  18. Application of genetic algorithms to in-core nuclear fuel management optimization

    International Nuclear Information System (INIS)

    Poon, P.W.; Parks, G.T.

    1993-01-01

    The search for an optimal arrangement of fresh and burnt fuel and control material within the core of a PWR represents a formidable optimization problem. The approach of combining the robust optimization capabilities of the Simulated Annealing (SA) algorithm with the computational speed of a Generalized Perturbation Theory (GPT) based evaluation methodology in the code FORMOSA has proved to be very effective. In this paper, we show that the incorporation of another stochastic search technique, a Genetic Algorithm, results in comparable optimization performance on serial computers and offers substantially superior performance on parallel machines. (orig.)

  19. Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

    International Nuclear Information System (INIS)

    Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

    2009-01-01

    The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

  20. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  1. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  2. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  3. Optimization of reload core design for PWR and application to Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shen Wei; Zhongsheng Xie; Banghua Yin

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the Linear Programming method using enrichments as control variable. In the second stage the optimum BP allocation is determined by the Flexible Tolerance Method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant(QNP)cycle 2 reloading design

  4. Out-of-core nuclear fuel cycle optimization utilizing an engineering workstation

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Comes, S.A.

    1986-01-01

    Within the past several years, rapid advances in computer technology have resulted in substantial increases in their performance. The net effect is that problems that could previously only be executed on mainframe computers can now be executed on micro- and minicomputers. The authors are interested in developing an engineering workstation for nuclear fuel management applications. An engineering workstation is defined as a microcomputer with enhanced graphics and communication capabilities. Current fuel management applications range from using workstations as front-end/back-end processors for mainframe computers to completing fuel management scoping calculations. More recently, interest in using workstations for final in-core design calculations has appeared. The authors have used the VAX 11/750 minicomputer, which is not truly an engineering workstation but has comparable performance, to complete both in-core and out-of-core fuel management scoping studies. In this paper, the authors concentrate on our out-of-core research. While much previous work in this area has dealt with decisions concerned with equilibrium cycles, the current project addresses the more realistic situation of nonequilibrium cycles

  5. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  6. Coarse-grained parallel genetic algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.; Lapa, Celso M.F.

    2003-01-01

    This work extends the research related to generic algorithms (GA) in core design optimization problems, which basic investigations were presented in previous work. Here we explore the use of the Island Genetic Algorithm (IGA), a coarse-grained parallel GA model, comparing its performance to that obtained by the application of a traditional non-parallel GA. The optimization problem consists on adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Our IGA implementation runs as a distributed application on a conventional local area network (LAN), avoiding the use of expensive parallel computers or architectures. After exhaustive experiments, taking more than 1500 h in 550 MHz personal computers, we have observed that the IGA provided gains not only in terms of computational time, but also in the optimization outcome. Besides, we have also realized that, for such kind of problem, which fitness evaluation is itself time consuming, the time overhead in the IGA, due to the communication in LANs, is practically imperceptible, leading to the conclusion that the use of expensive parallel computers or architecture can be avoided

  7. Out-of-core nuclear fuel cycle economic optimization for nonequilibrium cycles

    International Nuclear Information System (INIS)

    Comes, S.A.

    1987-01-01

    A methodology and associated computer code was developed to determine near-optimum out-of-core fuel management strategies. The code, named OCEON (Out-of-Core Economic OptimizationN), identified feed-region sizes and enrichments, and partially burned fuel-reload strategies for each cycle of a multi-cycle planning horizon, subject to cycle-energy requirements and constraints on feed enrichments, discharge burnups, and the moderator temperature coefficient. A zero-dimensional reactor physics model, enhanced by a linear reactivity model to provide batch power shares, performs the initial feed enrichment, burnup and constraint evaluations, while a two-dimensional, nodal code is used to refine the calculations for the final solutions. The economic calculations are performed rapidly using an annuity-factor-based model. Use of Monte Carlo integer programming to select the optimum solutions allows for the determination of a family of near-optimum solutions, from which engineering judgment may be used to select an appropriate strategy. Results from various nonequilibrium cycle energy requirement cases typically show a large number of low-cost solutions near the optimum. This confirms that the Monte Carlo integer programming approach of generating a family of solutions will be most useful for selecting optimum strategies when other considerations, such as incore loading pattern concerns, must be addressed

  8. Nuclear core baffling apparatus

    International Nuclear Information System (INIS)

    Cooper, F.W. Jr.; Silverblatt, B.L.; Knight, C.B.; Berringer, R.T.

    1979-01-01

    An apparatus for baffling the flow of reactor coolant fluid into and about the core of a nuclear reactor is described. The apparatus includes a plurality of longitudinally aligned baffle plates with mating surfaces that allow longitudinal growth with temperature increases while alleviating both leakage through the aligned plates and stresses on the components supporting the plates

  9. Nuclear core catchers

    International Nuclear Information System (INIS)

    Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

    1976-01-01

    A receptacle is described for taking the molten fragments of a nuclear reactor during a reactor core fusion accident. The receptacle is placed under the reactor. It includes at least one receptacle for the reactor core fragments, with a dome shaped part to distribute the molten fragments and at least one outside layer of alumina bricks around the dome. The characteristic of this receptacle is that the outer layer of bricks contains neutron poison rods which pass through the bricks and protrude in relation to them [fr

  10. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  11. The fuzzy clearing approach for a niching genetic algorithm applied to a nuclear reactor core design optimization problem

    International Nuclear Information System (INIS)

    Sacco, Wagner F.; Machado, Marcelo D.; Pereira, Claudio M.N.A.; Schirru, Roberto

    2004-01-01

    This article extends previous efforts on genetic algorithms (GAs) applied to a core design optimization problem. We introduce the application of a new Niching Genetic Algorithm (NGA) to this problem and compare its performance to these previous works. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. After exhaustive experiments we observed that our new niching method performs better than the conventional GA due to a greater exploration of the search space

  12. PWR in-core nuclear fuel management optimization utilizing nodal (non-linear NEM) generalized perturbation theory

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.; Kropaczek, D.J.

    1993-01-01

    The computational capability of efficiently and accurately evaluate reactor core attributes (i.e., k eff and power distributions as a function of cycle burnup) utilizing a second-order accurate advanced nodal Generalized Perturbation Theory (GPT) model has been developed. The GPT model is derived from the forward non-linear iterative Nodal Expansion Method (NEM) strategy, thereby extending its inherent savings in memory storage and high computational efficiency to also encompass GPT via the preservation of the finite-difference matrix structure. The above development was easily implemented into the existing coarse-mesh finite-difference GPT-based in-core fuel management optimization code FORMOSA-P, thus combining the proven robustness of its adaptive Simulated Annealing (SA) multiple-objective optimization algorithm with a high-fidelity NEM GPT neutronics model to produce a powerful computational tool used to generate families of near-optimum loading patterns for PWRs. (orig.)

  13. ESFR core optimization and uncertainty studies

    International Nuclear Information System (INIS)

    Rineiski, A.; Vezzoni, B.; Zhang, D.; Marchetti, M.; Gabrielli, F.; Maschek, W.; Chen, X.-N.; Buiron, L.; Krepel, J.; Sun, K.; Mikityuk, K.; Polidoro, F.; Rochman, D.; Koning, A.J.; DaCruz, D.F.; Tsige-Tamirat, H.; Sunderland, R.

    2015-01-01

    In the European Sodium Fast Reactor (ESFR) project supported by EURATOM in 2008-2012, a concept for a large 3600 MWth sodium-cooled fast reactor design was investigated. In particular, reference core designs with oxide and carbide fuel were optimized to improve their safety parameters. Uncertainties in these parameters were evaluated for the oxide option. Core modifications were performed first to reduce the sodium void reactivity effect. Introduction of a large sodium plenum with an absorber layer above the core and a lower axial fertile blanket improve the total sodium void effect appreciably, bringing it close to zero for a core with fresh fuel, in line with results obtained worldwide, while not influencing substantially other core physics parameters. Therefore an optimized configuration, CONF2, with a sodium plenum and a lower blanket was established first and used as a basis for further studies in view of deterioration of safety parameters during reactor operation. Further options to study were an inner fertile blanket, introduction of moderator pins, a smaller core height, special designs for pins, such as 'empty' pins, and subassemblies. These special designs were proposed to facilitate melted fuel relocation in order to avoid core re-criticality under severe accident conditions. In the paper further CONF2 modifications are compared in terms of safety and fuel balance. They may bring further improvements in safety, but their accurate assessment requires additional studies, including transient analyses. Uncertainty studies were performed by employing a so-called Total Monte-Carlo method, for which a large number of nuclear data files is produced for single isotopes and then used in Monte-Carlo calculations. The uncertainties for the criticality, sodium void and Doppler effects, effective delayed neutron fraction due to uncertainties in basic nuclear data were assessed for an ESFR core. They prove applicability of the available nuclear data for ESFR

  14. Nuclear reactor core stabilizing arrangement

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

  15. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  16. Nuclear reactor core

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo; Ishibashi, Yoko; Mochida, Takaaki; Haikawa, Katsumasa; Yamanaka, Akihiro.

    1995-01-01

    A reactor core is radially divided into an inner region, an outer region and an outermost region. As a fuel, three kinds of fuels, namely, a high enrichment degree fuel at 3.4%, a middle enrichment degree fuel at 2.3% and a low enrichment degree at 1.1% of a fuel average enrichment degree of fission product are used. Each of the fuels is bisected to upper and lower portions at an axial center thereof. The difference of average enrichment degrees between upper and lower portions is 0.1% for the high enrichment degree fuel, 0.3% for the middle enrichment degree fuel and 0.2% for the low enrichment degree fuel. In addition, the composition of fuels in each of radial regions comprises 100% of the low enrichment degree fuels in the outermost region, 91% of the higher enrichment degree fuels and 9% of the middle enrichment degree fuels in the outer region, and 34% of the high enrichment degree fuels and 30% of the middle enrichment degree fuels in the inner region. With such a constitution, fuel economy can be improved while maintaining the thermal margin in an initially loaded reactor core of a BWR type reactor. (I.N.)

  17. Nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F

    1974-07-11

    The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

  18. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  19. Nuclear reactor core safety device

    International Nuclear Information System (INIS)

    Colgate, S.A.

    1977-01-01

    The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

  20. Nuclear reactor core servicing apparatus

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved core servicing apparatus for a nuclear reactor of the type having a reactor vessel, a vessel head having a head penetration therethrough, a removable plug adapted to fit in the head penetration, and a core of the type having an array of elongated assemblies. The improved core servicing apparatus comprises a plurality of support columns suspended from the removable plug and extending downward toward the nuclear core, rigid support means carried by each of the support columns, and a plurality of servicing means for each of the support columns for servicing a plurality of assemblies. Each of the plurality of servicing means for each of the support columns is fixedly supported in a fixed array from the rigid support means. Means are provided for rotating the rigid support means and servicing means between condensed and expanded positions. When in the condensed position, the rigid support means and servicing means lie completely within the coextensive boundaries of the plug, and when in the expanded position, some of the rigid support means and servicing means lie without the coextensive boundaries of the plug

  1. Hyper-heuristic applied to nuclear reactor core design

    International Nuclear Information System (INIS)

    Domingos, R P; Platt, G M

    2013-01-01

    The design of nuclear reactors gives rises to a series of optimization problems because of the need for high efficiency, availability and maintenance of security levels. Gradient-based techniques and linear programming have been applied, as well as genetic algorithms and particle swarm optimization. The nonlinearity, multimodality and lack of knowledge about the problem domain makes de choice of suitable meta-heuristic models particularly challenging. In this work we solve the optimization problem of a nuclear reactor core design through the application of an optimal sequence of meta-heuritics created automatically. This combinatorial optimization model is known as hyper-heuristic.

  2. Methods and techniques of nuclear in-core fuel management

    International Nuclear Information System (INIS)

    Jong, A.J. de.

    1992-04-01

    Review of methods of nuclear in-core fuel management (the minimal critical mass problem, minimal power peaking) and calculational techniques: reactorphysical calculations (point reactivity models, continuous refueling, empirical methods, depletion perturbation theory, nodal computer programs); optimization techniques (stochastic search, linear programming, heuristic parameter optimization). (orig./HP)

  3. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  4. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  5. Simulated annealing algorithm for reactor in-core design optimizations

    International Nuclear Information System (INIS)

    Zhong Wenfa; Zhou Quan; Zhong Zhaopeng

    2001-01-01

    A nuclear reactor must be optimized for in core fuel management to make full use of the fuel, to reduce the operation cost and to flatten the power distribution reasonably. The author presents a simulated annealing algorithm. The optimized objective function and the punishment function were provided for optimizing the reactor physics design. The punishment function was used to practice the simulated annealing algorithm. The practical design of the NHR-200 was calculated. The results show that the K eff can be increased by 2.5% and the power distribution can be flattened

  6. Nuclear reactor with several cores

    International Nuclear Information System (INIS)

    Swars, H.

    1977-01-01

    Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

  7. A novel optimization method, Gravitational Search Algorithm (GSA), for PWR core optimization

    International Nuclear Information System (INIS)

    Mahmoudi, S.M.; Aghaie, M.; Bahonar, M.; Poursalehi, N.

    2016-01-01

    Highlights: • The Gravitational Search Algorithm (GSA) is introduced. • The advantage of GSA is verified in Shekel’s Foxholes. • Reload optimizing in WWER-1000 and WWER-440 cases are performed. • Maximizing K eff , minimizing PPFs and flattening power density is considered. - Abstract: In-core fuel management optimization (ICFMO) is one of the most challenging concepts of nuclear engineering. In recent decades several meta-heuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. This paper presents a new method of using Gravitational Search Algorithm (GSA) for in-core fuel management optimization. The GSA is constructed based on the law of gravity and the notion of mass interactions. It uses the theory of Newtonian physics and searcher agents are the collection of masses. In this work, at the first step, GSA method is compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step for finding the best core, the GSA algorithm has been performed for three PWR test cases including WWER-1000 and WWER-440 reactors. In these cases, Multi objective optimizations with the following goals are considered, increment of multiplication factor (K eff ), decrement of power peaking factor (PPF) and power density flattening. It is notable that for neutronic calculation, PARCS (Purdue Advanced Reactor Core Simulator) code is used. The results demonstrate that GSA algorithm have promising performance and could be proposed for other optimization problems of nuclear engineering field.

  8. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    Brown, W.L.; Geronime, R.L.

    1978-01-01

    Sensors including radiation detectors and the like for use within the core of nuclear reactors and which are constructed in a manner to provide optimum reliability of the sensor during use are described

  9. Method for refuelling a nuclear reactor core

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This invention relates to an improved method for refuelling a nuclear reactor core inside a reactor vessel. The technique allows a substantial reduction in the refuelling time as compared with previously known methods and permits fewer out of core operations and smaller temporary storage space. (U.K.)

  10. Perturbation theory in nuclear fuel management optimization

    International Nuclear Information System (INIS)

    Ho, L.W.

    1981-01-01

    Nuclear in-core fuel management involves all the physical aspects which allow optimal operation of the nuclear fuel within the reactor core. In most nuclear power reactors, fuel loading patterns which have a minimum power peak are economically desirable to allow the reactors to operate at the highest power density and to minimize the possibility of fuel failure. In this study, perturbation theory along with a binary fuel shuffling technique is applied to predict the effects of various core configurations, and hence, the optimization of in-core fuel management. The computer code FULMNT has been developed to shuffle the fuel assemblies in search of the lowest possible power peaking factor. An iteration approach is used in the search routine. A two-group diffusion theory method is used to obtain the power distribution for the iterations. A comparison of the results of this method with other methods shows that this approach can save computer time. The code also has a burnup capability which can be used to check power peaking throughout the core life

  11. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  12. Core baffle for nuclear reactors

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1977-01-01

    The invention concerns the design of the core of a LWR with a large number of fuel assemblies formed by fuel rods and kept in position by spacer grids. According to the invention, at the level of the spacer grids match plates are mounted with openings so the flow of coolant directed upwards will not be obstructed and a parallel bypass will be obtained in the space between the core barrel and the baffle plates. In case of an accident, this configuration reduces or avoids damage from overpressure reactions. (HP) [de

  13. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  14. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1979-01-01

    A lateral restraint and control systemm for a nuclear reactor core provides an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit is composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased by an amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  15. Magnetic nuclear core restraint and control

    International Nuclear Information System (INIS)

    Cooper, M.H.

    1978-01-01

    Disclosed is a lateral restraint and control system for a nuclear reactor core adaptable to provide an inherent decrease of core reactivity in response to abnormally high reactor coolant fluid temperatures. An electromagnet is associated with structure for radially compressing the core during normal reactor conditions. A portion of the structures forming a magnetic circuit are composed of ferromagnetic material having a curie temperature corresponding to a selected coolant fluid temperature. Upon a selected signal, or inherently upon a preselected rise in coolant temperature, the magnetic force is decreased a given amount sufficient to relieve the compression force so as to allow core radial expansion. The expanded core configuration provides a decreased reactivity, tending to shut down the nuclear reaction

  16. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  17. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  18. Core optimization studies at JEN-Spain

    International Nuclear Information System (INIS)

    Gomez Alonso, M.

    1983-01-01

    The JEN-1 is a 3-MW reactor which uses flat-plate fuel elements. It was originally fueled with 20%-enriched uranium but more recently with 90%-enriched fuel. It now appears that it will have to be converted back to using 20%- enriched fuel. Progress is presently being made in fuel fabrication. Plates with meat thicknesses of up to 1.5 mm have been fabricated. Plates are being tested with 40 wt % uranium in the fuel meat. Progress is also being made in reactor design in collaboration with atomic energy commissions of other countries for swimming pool reactors being designed or under construction in Chile, Ecuador, and Spain itself. The design studies address core optimization, safety analysis report updating, irradiation facilities, etc. Core optimization is specifically addressed in this paper. A common swimming-pool-type reactor such as the JEN-1 served as an example. The philosophy adopted in this study is not to try to match the high enrichment core, but rather to treat the design as new and try to optimize it using simplified neutronic/thermal hydraulic/economic models. This philosophy appears to be somewhat original. As many as possible of the fuel parameters are constrained to remain constant

  19. Core construction for nuclear reactors

    International Nuclear Information System (INIS)

    Pettinger, D.S.

    1977-01-01

    HTR core construction with prismatic graphite blocks piled into columns. The front of the blocks is concavely curved. The lines of contact of two blocks are always not vertical, i.e. the blocks of one column are supported by the blocks of neighbouring columns so that ducts are formed. Groups of three or four of these columns may additionally be arranged around a central column which has recesses in order to lock the blocks of one group together. With this arrangement, dimensional changes of the graphite blocks under operating conditions can be taken up. (DG) [de

  20. Nuclear fuel management optimization for LWRs

    International Nuclear Information System (INIS)

    Turinsky, Paul J.

    1997-01-01

    LWR in core nuclear fuel management involves the placement of fuel and control materials so that a specified objective is achieved within constraints. Specifically, one is interested in determining the core loading pattern (LP of fuel assemblies and burnable poisons and for BWR, also control rod insertion versus cycle exposure. Possible objectives include minimization of feed enrichment and maximization of cycle energy production, discharge burnup or thermal margin. Constraints imposed relate to physical constraints, e.g. no discrete burnable poisons in control rod locations, and operational and safety constraints, e.g. maximum power peaking limit. The LP optimization problem is a large scale, nonlinear, mixed-integer decision variables problem with active constraints. Even with quarter core symmetry imposed, there are above 10 100 possible LPs. The implication is that deterministic optimization methods are not suitable, so in this work we have pursued using the stochastic Simulated Annealing optimization method. Adaptive penalty functions are used to impose certain constraints, allowing unfeasible regions of the search space to be transverse. Since ten of thousands of LPs must be examined to achieve high computational efficiency, higher-order Generalized Perturbation Theory is utilized to solve the Nodal Expansion Method for of the two-group neutron diffusion. These methods have been incorporated into the FORMOSA series of codes and used to optimize PWR and BWR reload cores. (author). 9 refs., 3 tabs

  1. Nuclear clustering - a cluster core model study

    International Nuclear Information System (INIS)

    Paul Selvi, G.; Nandhini, N.; Balasubramaniam, M.

    2015-01-01

    Nuclear clustering, similar to other clustering phenomenon in nature is a much warranted study, since it would help us in understanding the nature of binding of the nucleons inside the nucleus, closed shell behaviour when the system is highly deformed, dynamics and structure at extremes. Several models account for the clustering phenomenon of nuclei. We present in this work, a cluster core model study of nuclear clustering in light mass nuclei

  2. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  3. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  4. Apparatus for controlling nuclear core debris

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

  5. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  6. Core support structure for nuclear power plants

    International Nuclear Information System (INIS)

    Steinkamp, E.; Tautz, J.; Ries, H.

    1979-01-01

    A core support structure for nuclear power plants includes a grid of mutually crossing bridges and a support ring surrounding the grid and connected to ends of the outer bridges of the grid, the grid being formed of profile rod crosses having legs of given length, respective legs of pairs of adjacent crosses abutting one another endwise to form together a side of the smallest mesh opening of the grid, and weld means for securing the profile rod crosses to one another at the mutually abutting ends of the legs thereof; and method of producing the foregoing core support structure

  7. Optimal processor for malfunction detection in operating nuclear reactor

    International Nuclear Information System (INIS)

    Ciftcioglu, O.

    1990-01-01

    An optimal processor for diagnosing operational transients in a nuclear reactor is described. Basic design of the processor involves real-time processing of noise signal obtained from a particular in core sensor and the optimality is based on minimum alarm failure in contrast to minimum false alarm criterion from the safe and reliable plant operation viewpoint

  8. A loading pattern optimization method for nuclear fuel management

    International Nuclear Information System (INIS)

    Argaud, J.P.

    1997-01-01

    Nuclear fuel reload of PWR core leads to the search of an optimal nuclear fuel assemblies distribution, namely of loading pattern. This large discrete optimization problem is here expressed as a cost function minimization. To deal with this problem, an approach based on gradient information is used to direct the search in the patterns discrete space. A method using an adjoint state formulation is then developed, and final results of complete patterns search tests by this method are presented. (author)

  9. Engineering fuel reloading sequence optimization for in-core shuffling system

    International Nuclear Information System (INIS)

    Jeong, Seo G.; Suh, Kune Y.

    2008-01-01

    Optimizing the nuclear fuel reloading process is central to enhancing the economics of nuclear power plant (NPP). There are two kinds of reloading method: in-core shuffling and ex-core shuffling. In-core shuffling has an advantage of reloading time when compared with ex-core shuffling. It is, however, not easy to adopt an in-core shuffling because of additional facilities required and regulations involved at the moment. The in-core shuffling necessitates minimizing the movement of refueling machine because reloading paths can be varied according to differing reloading sequences. In the past, the reloading process depended on the expert's knowledge and experience. Recent advances in computer technology have apparently facilitated the heuristic approach to nuclear fuel reloading sequence optimization. This work presents a first in its kind of in-core shuffling whereas all the Korean NPPs have so far adopted ex-core shuffling method. Several plants recently applied the in-core shuffling strategy, thereby saving approximately 24 to 48 hours of outage time. In case of in-core shuffling one need minimize the movement of refueling machine because reloading path can be varied according to different reloading sequences. Advances in computer technology have enabled optimizing the in-core shuffling by solving a traveling salesman problem. To solve this problem, heuristic algorithm is used, such as ant colony algorithm and genetic algorithm. The Systemic Engineering Reload Analysis (SERA) program is written to optimize shuffling sequence based on heuristic algorithms. SERA is applied to the Optimized Power Reactor 1000 MWe (OPR1000) on the assumption that the NPP adopts the in-core shuffling in the foreseeable future. It is shown that the optimized shuffling sequence resulted in reduced reloading time. (author)

  10. A supercomputing application for reactors core design and optimization

    International Nuclear Information System (INIS)

    Hourcade, Edouard; Gaudier, Fabrice; Arnaud, Gilles; Funtowiez, David; Ammar, Karim

    2010-01-01

    Advanced nuclear reactor designs are often intuition-driven processes where designers first develop or use simplified simulation tools for each physical phenomenon involved. Through the project development, complexity in each discipline increases and implementation of chaining/coupling capabilities adapted to supercomputing optimization process are often postponed to a further step so that task gets increasingly challenging. In the context of renewal in reactor designs, project of first realization are often run in parallel with advanced design although very dependant on final options. As a consequence, the development of tools to globally assess/optimize reactor core features, with the on-going design methods accuracy, is needed. This should be possible within reasonable simulation time and without advanced computer skills needed at project management scale. Also, these tools should be ready to easily cope with modeling progresses in each discipline through project life-time. An early stage development of multi-physics package adapted to supercomputing is presented. The URANIE platform, developed at CEA and based on the Data Analysis Framework ROOT, is very well adapted to this approach. It allows diversified sampling techniques (SRS, LHS, qMC), fitting tools (neuronal networks...) and optimization techniques (genetic algorithm). Also data-base management and visualization are made very easy. In this paper, we'll present the various implementing steps of this core physics tool where neutronics, thermo-hydraulics, and fuel mechanics codes are run simultaneously. A relevant example of optimization of nuclear reactor safety characteristics will be presented. Also, flexibility of URANIE tool will be illustrated with the presentation of several approaches to improve Pareto front quality. (author)

  11. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    Marth, W.; Schroeder, R.

    1983-07-01

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.) [de

  12. Gas core nuclear rocket feasibility project

    International Nuclear Information System (INIS)

    Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1997-09-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas core nuclear rocket (GCNR) has the potential to be such a system. The gas core concept relies on the use of fluid dynamic forces to create and maintain a vortex. The vortex is composed of a fissile material which will achieve criticality and produce high power levels. By radiatively coupling to the surrounding fluids, extremely high temperatures in the propellant and, thus, high specific impulses can be generated. The ship velocities enabled by such performance may allow a 9 month round trip, manned Mars mission to be considered. Alternatively, one might consider slightly longer missions in ships that are heavily shielded against the intense Galactic Cosmic Ray flux to further reduce the radiation dose to the crew. The current status of the research program at the Los Alamos National Laboratory into the gas core nuclear rocket feasibility will be discussed

  13. Nuclear reactor core and fuel element therefor

    International Nuclear Information System (INIS)

    Fortescue, P.

    1986-01-01

    This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces

  14. Sensors for use in nuclear reactor cores

    International Nuclear Information System (INIS)

    1980-01-01

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter. (UK)

  15. Sensors for use in nuclear reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-21

    A neutron sensor is described for use in nuclear reactor cores which does not require external power but merely an emitter, a collector and an insulator material between the two to generate an electric current that is indicative of the intensity of the radiation. The sensor is manufactured in such a way that brazed joints or spices are avoided and the insulation material used may be of relatively low density of compaction and will center the emitter and the lead wire with respect to the outer sheath or tube without deformation or varying geometry of the center wire or emitter.

  16. Core of a fast neutron nuclear reactor

    International Nuclear Information System (INIS)

    Giacometti, Christian; Mougniot, J.-C.; Ravier, Jean.

    1974-01-01

    The fast neutron nuclear reactor described includes an internal area in fissile material completely enclosed in an area of fertile material forming the outside blanket. The internal fissile area is provided with housings exclusively filled with fertile material forming one or more inside blankets. In this core the internal blankets are shaped like rings vertically separating superimposed rings of fissile material. The blanket of material nearest to the periphery is circumscribed externally by a contour having an indented shape on its straight section so as to increase the contact area between this blanket and the external blanket [fr

  17. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  18. Core Technology Development of Nuclear spin polarization

    International Nuclear Information System (INIS)

    Yoo, Byung Duk; Gwon, Sung Ok; Kwon, Duck Hee; Lee, Sung Man

    2009-12-01

    In order to study nuclear spin polarization, we need several core technologies such as laser beam source to polarize the nuclear spin, low pressured helium cell development whose surface is essential to maintain polarization otherwise most of the polarized helium relaxed in short time, development of uniform magnetic field system which is essential for reducing relaxation, efficient vacuum system, development of polarization measuring system, and development of pressure raising system about 1000 times. The purpose of this study is to develop resonable power of laser system, that is at least 5 watt, 1083 nm, 4GHz tuneable. But the limitation of this research fund enforce to develop amplifying system into 5 watt with 1 watt system utilizing laser-diod which is already we have in stock. We succeeded in getting excellent specification of fiber laser system with power of 5 watts, 2 GHz linewidth, more than 80 GHz tuneable

  19. Core access system for nuclear reactor

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

  20. Fuel assembly and nuclear reactor core

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Aoyama, Motoo; Yamashita, Jun-ichi.

    1995-01-01

    The present invention concerns a fuel assembly and a nuclear reactor core capable of improving a transmutation rate of transuranium elements while improving a residual rate of fission products. In a reactor core of a BWR type reactor to which fuel rods with transuranium elements (TRU) enriched are loaded, the enrichment degree of transuranium elements occupying in fuel materials is determined not less than 2wt%, as well as a ratio of number of atoms between hydrogen and fuel heavy metals in an average reactor core under usual operation state (H/HM) is determined not more than 3 times. In addition, a ratio of the volumes between coolant regions and fuel material regions is determined not more than 2 times. A T ratio (TRU/Pu) is lowered as the TRU enrichment degree is higher and the H/HM ratio is lower. In order to reduce the T ratio not more than 1, the TRU enrichment degree is determined as not less than 2wt%, and the H/HM ratio is determined to not more than 3 times. Accordingly, since the H/HM ratio is reduced to not more than 1, and TRU is transmuted while recycling it with plutonium, the transmutation ratio of transuranium elements can be improved while improving the residual rate of fission products. (N.H.)

  1. The development of direct core monitoring in Nuclear Electric plc

    International Nuclear Information System (INIS)

    Curtis, R.F.; Jones, S. Reed, J.; Wickham, A.J.

    1996-01-01

    Monitoring of graphite behaviour in Nuclear Electric Magnox and AGR reactors is necessary to support operating safety cases and to ensure that reactor operation is optimized to sustain the necessary core integrity for the economic life of the reactors. The monitoring programme combines studies for pre-characterized ''installed'' samples with studies on samples trepanned from within the cores and also with studies of core and channel geometry using specially designed equipment. Nuclear Electric has two trepanning machines originally designed for Magnox-reactor work which have been used for a substantial programme over many years. They have recently been upgraded to improve sampling speed, safety and versatility - the last being demonstrated by their adaptation for a recently-won contract associated with decommissioning the Windscale piles. Radiological hazards perceived when the AGR trepanning system was designed resulted in very cumbersome equipment. This has worked well but has been inconvenient in operation. The development of a smaller and improved system for deploying the equipment is now reported. Channel dimension monitoring equipment is discussed in detail with examples of data recovered from both Magnox and AGR cores. A resolution of ± 2 of arc (tilt) and ± 0.01 mm change in diameter in attainable. It is also theoretically possible to establish brick stresses by measuring geometry changes which result from trepanning. Current development work on a revolving scanning laser rangefinder which will enable the measurement of diameters to a resolution of 0.001 mm will also be discussed. This paper also discusses non-destructive techniques for crack detection employing ultrasound or resistance networks, the use of special manipulators to deliver inspection and repair equipment and recent developments to install displacement monitors in peripheral regions of the cores, to aid the understanding of the interaction of the restraint system with the core - the region

  2. SMART core preliminary nuclear design-II

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Chan; Ji, Seong Kyun; Chang, Moon Hee

    1997-06-01

    Three loading patterns for 330 MWth SMART core are constructed for 25, 33 and 29 CRDMs, and one loading pattern for larger 69-FA core with 45 CRDMs is also constructed for comparison purpose. In this study, the core consists of 57 reduced height Korean Optimized Fuel Assemblies (KOFAs) developed by KAERI. The enrichment of fuel is 4.95 w/o. As a main burnable poison, 35% B-10 enriched B{sub 4}C-Al{sub 2}O{sub 3} shim is used. To control stuck rod worth, some gadolinia bearing fuel rods are used. The U-235 enrichment of the gadolinia bearing fuel rods is 1.8 w/o as used in KOFA. All patterns return cycle length of about 3 years. Three loading patterns except 25-CRDM pattern satisfy cold shutdown condition of keff {<=} 0.99 without soluble boron. These three patterns also satisfy the refueling condition of keff {<=} 0.95. In addition to the construction of loading pattern, an editing module of MASTER PPI files for rod power history generation is developed and rod power histories are generated for 29-CRDM loading pattern. Preliminary Fq design limit is suggested as 3.71 based on KOFA design experience. (author). 9 tabs., 45 figs., 16 refs.

  3. Optimization of burnable poison disposition for in-core fuel assemblies

    International Nuclear Information System (INIS)

    Zhong Wenfa; Luo Rong; Zhou Quan

    1997-09-01

    The optimization of the burnable poison disposition in the initial core loading of the 200 MW nuclear heating reactor (NHR-200), is studied. The mass fraction of the burnable poison is used as the control variable with the objective to minimize the power peaking factor. The flexible tolerance method is used to solve the nonlinear programming optimal problem. The optimization method can be used in reactor physics design, and get a new pattern of initial core which is of reference value. (2 refs., 8 figs., 1 tab.)

  4. Preliminary analysis on in-core fuel management optimization of molten salt pebble-bed reactor

    International Nuclear Information System (INIS)

    Xia Bing; Jing Xingqing; Xu Xiaolin; Lv Yingzhong

    2013-01-01

    The Nuclear Hot Spring (NHS) is a molten salt pebble-bed reactor featured by full power natural circulation. The unique horizontal coolant flow of the NHS demands the fuel recycling schemes based on radial zoning refueling and the corresponding method of fuel management optimization. The local searching algorithm (LSA) and the simulated annealing algorithm (SAA), the stochastic optimization methods widely used in the refueling optimization problems in LWRs, were applied to the analysis of refueling optimization of the NHS. The analysis results indicate that, compared with the LSA, the SAA can survive the traps of local optimized solutions and reach the global optimized solution, and the quality of optimization of the SAA is independent of the choice of the initial solution. The optimization result gives excellent effects on the in-core power flattening and the suppression of fuel center temperature. For the one-dimensional zoning refueling schemes of the NHS, the SAA is an appropriate optimization method. (authors)

  5. Quantum behaved Particle Swarm Optimization with Differential Mutation operator applied to WWER-1000 in-core fuel management optimization

    International Nuclear Information System (INIS)

    Jamalipour, Mostafa; Sayareh, Reza; Gharib, Morteza; Khoshahval, Farrokh; Karimi, Mahmood Reza

    2013-01-01

    Highlights: ► A new method called QPSO-DM is applied to BNPP in-core fuel management optimization. ► It is found that QPSO-DM performs better than PSO and QPSO. ► This method provides a permissible arrangement for optimum loading pattern. - Abstract: This paper presents a new method using Quantum Particle Swarm Optimization with Differential Mutation operator (QPSO-DM) for optimizing WWER-1000 core fuel management. Genetic Algorithm (GA) and Particle Swarm Optimization (PSO) have shown good performance on in-core fuel management optimization (ICFMO). The objective of this paper is to show that QPSO-DM performs very well and is comparable to PSO and Quantum Particle Swarm Optimization (QPSO). Most of the strategies for ICFMO are based on maximizing multiplication factor (k eff ) to increase cycle length and minimizing power peaking factor (P q ) in order to improve fuel integrity. PSO, QPSO and QPSO-DM have been implemented to fulfill these requirements for the first operating cycle of WWER-1000 Bushehr Nuclear Power Plant (BNPP). The results show that QPSO-DM performs better than the others. A program has been written in MATLAB to map PSO, QPSO and QPSO-DM for loading pattern optimization. WIMS and CITATION have been used to simulate reactor core for neutronic calculations

  6. Genetic algorithms applied to nuclear reactor design optimization

    International Nuclear Information System (INIS)

    Pereira, C.M.N.A.; Schirru, R.; Martinez, A.S.

    2000-01-01

    A genetic algorithm is a powerful search technique that simulates natural evolution in order to fit a population of computational structures to the solution of an optimization problem. This technique presents several advantages over classical ones such as linear programming based techniques, often used in nuclear engineering optimization problems. However, genetic algorithms demand some extra computational cost. Nowadays, due to the fast computers available, the use of genetic algorithms has increased and its practical application has become a reality. In nuclear engineering there are many difficult optimization problems related to nuclear reactor design. Genetic algorithm is a suitable technique to face such kind of problems. This chapter presents applications of genetic algorithms for nuclear reactor core design optimization. A genetic algorithm has been designed to optimize the nuclear reactor cell parameters, such as array pitch, isotopic enrichment, dimensions and cells materials. Some advantages of this genetic algorithm implementation over a classical method based on linear programming are revealed through the application of both techniques to a simple optimization problem. In order to emphasize the suitability of genetic algorithms for design optimization, the technique was successfully applied to a more complex problem, where the classical method is not suitable. Results and comments about the applications are also presented. (orig.)

  7. Core design and operation optimization methods based on time-dependent perturbation theory

    International Nuclear Information System (INIS)

    Greenspan, E.

    1983-08-01

    A general approach for the optimization of nuclear reactor core design and operation is outlined; it is based on two cornerstones: a newly developed time-dependent (or burnup-dependent) perturbation theory for nonlinear problems and a succesive iteration technique. The resulting approach is capable of handling realistic reactor models using computational methods of any degree of sophistication desired, while accounting for all the constraints imposed. Three general optimization strategies, different in the way for handling the constraints, are formulated. (author)

  8. Application of a genetic algorithm to core reload pattern optimization

    International Nuclear Information System (INIS)

    Tanker, E.; Tanker, A.Z.

    1994-01-01

    A genetic algorithm is applied to reload pattern optimization of a PWR core. Evaluating all different distributions of a given batch load separately is found slow and ineffective. Allowing patterns from different distributions to combine reproduce, an optimized pattern better than that obtained from from linear programming is found, albeit in a longer time. (authors). 5 refs., 2 tabs

  9. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  10. Optimal core acquisition and remanufacturing policies under uncertain core quality fractions

    NARCIS (Netherlands)

    Teunter, R.H.; Flapper, S.D.P.

    2011-01-01

    Cores acquired by a remanufacturer are typically highly variable in quality. Even if the expected fractions of the various quality levels are known, then the exact fractions when acquiring cores are still uncertain. Our model incorporates this uncertainty in determining optimal acquisition decisions

  11. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

    In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

  12. Reloading optimization of pressurized water reactor core with burnable absorber fuel

    International Nuclear Information System (INIS)

    Shi Xiuan; Liu Zhihong; Hu Yongming

    2008-01-01

    The reloading optimization problem of PWR with burnable absorber fuel is very difficult, and common optimization algorithms are inefficient and have bad global performance for it. Characteristic statistic algorithm (CSA) is very fit for the problem. In the past, the reloading optimization using CSA has shortcomings of separating the fuel assemblies' loading pattern (LP) optimization from burnable absorber's placement (BP) optimization. In this study, LP and BP were optimized simultaneously using CSA coupled with CYCLE2D, which is a core analysis code. The corresponding reloading coupling optimization software, CSALPBP, was developed. The 10th cycle reloading design of Daya Bay Nuclear Power Plant was optimized using CSALPBP. The results show that CSALPBP has high efficiency and excellent global performance. (authors)

  13. Fast three-dimensional core optimization based on modified one-group model

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Fernando S. [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Rio de Janeiro, RJ (Brazil). Dept. GCN-T], e-mail: freire@eletronuclear.gov.br; Martinez, Aquilino S.; Silva, Fernando C. da [Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear], e-mail: aquilino@con.ufrj.br, e-mail: fernando@con.ufrj.br

    2009-07-01

    The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

  14. Fast three-dimensional core optimization based on modified one-group model

    International Nuclear Information System (INIS)

    Freire, Fernando S.; Martinez, Aquilino S.; Silva, Fernando C. da

    2009-01-01

    The optimization of any nuclear reactor core is an extremely complex process that consumes a large amount of computer time. Fortunately, the nuclear designer can rely on a variety of methodologies able to approximate the analysis of each available core loading pattern. Two-dimensional codes are usually used to analyze the loading scheme. However, when particular axial effects are present in the core, two-dimensional analysis cannot produce good results and three-dimensional analysis can be required at all time. Basically, in this paper are presented the major advantages that can be found when one use the modified one-group diffusion theory coupled with a buckling correction model in optimization process. The results of the proposed model are very accurate when compared to benchmark results obtained from detailed calculations using three-dimensional nodal codes (author)

  15. Development of a VVER-1000 core loading pattern optimization program based on perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2012-01-01

    Highlights: ► We use perturbation theory to find an optimum fuel loading pattern in a VVER-1000. ► We provide a software for in-core fuel management optimization. ► We consider two objectives for our method (perturbation theory). ► We show that perturbation theory method is very fast and accurate for optimization. - Abstract: In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1 days, as well.

  16. Modeling of the core of Atucha II nuclear power plant

    International Nuclear Information System (INIS)

    Blanco, Anibal

    2007-01-01

    This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

  17. Automated Design and Optimization of Pebble-bed Reactor Cores

    International Nuclear Information System (INIS)

    Gougar, Hans D.; Ougouag, Abderrafi M.; Terry, William K.

    2010-01-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  18. Core clamping device for a nuclear reactor

    International Nuclear Information System (INIS)

    Guenther, R.W.

    1974-01-01

    The core clamping device for a fast neutron reactor includes clamps to support the fuel zone against the pressure vessel. The clamps are arranged around the circumference of the core. They consist of torsion bars arranged parallel at some distance around the core with lever arms attached to the ends whose force is directed in the opposite direction, pressing against the wall of the pressure vessel. The lever arms and pressure plates also actuated by the ends of the torsion bars transfer the stress, the pressure plates acting upon the fuel elements or fuel assemblies. Coupling between the ends of the torsion bars and the pressure plates is achieved by end carrier plates directly attached to the torsion bars and radially movable. This clamping device follows the thermal expansions of the core, allows specific elements to be disengaged in sections and saves space between the core and the neutron reflectors. (DG) [de

  19. Optimized core loading sequence for Ukraine WWER-1000 reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Fuel Assemblies (WFAs) experienced mechanical damage of the grids during loading at both South Ukraine 2 (SU2) and South Ukraine 3 (SU3). The grids were damaged due to high lateral loads exceeding their strength limit. The high lateral loads were caused by a combination of distortion and stiffness of the mixed core fuel assemblies and significant fuel assembly-to-fuel assembly interaction combined with the core loading sequence being used. To prevent damage of the WFA grids during core loading, Westinghouse has developed a loading sequence technique and loading aides (smooth sided dummies and top nozzle loading guides) designed to minimize fuel assembly-to-fuel assembly interaction while maximizing the potential for successful loading (i.e., no fuel assembly damage and minimized loading time). The loading sequence technique accounts for cycle-specific core loading patterns and is based on previous Westinghouse WWER core loading experience and fundamental principles. The loading aids are developed to “open-up” the target core location or to provide guidance into a target core location. The Westinghouse optimized core loading sequence and smooth sided dummies were utilized during the successful loading of SU3 Cycle 25 mixed core in March 2015, with no instances of fuel assembly damage and yet still provided considerable time savings relative to the 2012 and 2013 SU3 reload campaigns. (authors)

  20. Optimal burnable poison utilization in PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

  1. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  2. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    preferred working fluid and the HP working temperature can be as high as 1300 K. It is feasible to achieve criticality and to maintain a nearly zero burn-up reactivity swing for at least 20 EFPY with an average linear heat generation rate (LHR) of 90W/cm. The preferred design utilizes nitride fuel made of natural nitrogen and loaded with depleted uranium and TRU from LWR spent fuel cooled for approximately 30 years. The preferred intermediate coolant is LiF BeF 2 ; its average outlet temperature is ∼ 1040K. Effective heat transfer to the intermediate coolant is obtained with HPs extending out of the core less than 50 cm. The required reactor vessel height is significantly smaller than that of the reference ENHS: 9 vs. ∼20 m. The vessel diameter is slightly larger: 4 vs. ∼ 3.5 m. In conclusion, it appears feasible to design a HP-ENHS reactor to achieve its primary design objectives. The resulting HP-ENHS reactor concept is unique in offering sustainable proliferation-resistant nuclear energy that can be delivered at very high temperatures. A number of outstanding issues need be addressed, though, before the practicality of the HP design concept could be asserted. Included among these issues are: (1) More thorough reactor safety analysis, including transient analysis; (2) Fuel-cladding chemical compatibility; (3) Manufacturability and welding of Mo TZM alloy; (4) Maximization of the specific power by optimization of fuel/HP diameter and core length; and (5) Economic analysis

  3. Neutronic analysis of the ford nuclear reactor leu core

    International Nuclear Information System (INIS)

    Raza, S.S.; Hayat, T.

    1989-08-01

    Neutronic analysis of the ford nuclear reactor low enriched uranium core has been carried out to gain confidence in the com puting methodology being used for Pakistan Research Reactor-1 core conversion calculations. The computed value of the effective multiplication factor (Keff) is found to be in good agreement with that quoted by others. (author). 6 figs

  4. Supporting system for the core restraint of nuclear reactors

    International Nuclear Information System (INIS)

    Kaser, A.

    1973-01-01

    The core restraint of water cooled nuclear reactors which is needed to direct the flow of the coolant through the core can be manufactured only in a moderate wall thickness. Thus, the majority of the loads have to be transmitted to the core barrel which is more rigid. The patent refers to a system of circumferential and vertical support members most of which are free to move relatively to each other, thus reducing thermal stresses during operation. (P.K.)

  5. Optimizations of Unstructured Aerodynamics Computations for Many-core Architectures

    KAUST Repository

    Al Farhan, Mohammed Ahmed

    2018-04-13

    We investigate several state-of-the-practice shared-memory optimization techniques applied to key routines of an unstructured computational aerodynamics application with irregular memory accesses. We illustrate for the Intel KNL processor, as a representative of the processors in contemporary leading supercomputers, identifying and addressing performance challenges without compromising the floating point numerics of the original code. We employ low and high-level architecture-specific code optimizations involving thread and data-level parallelism. Our approach is based upon a multi-level hierarchical distribution of work and data across both the threads and the SIMD units within every hardware core. On a 64-core KNL chip, we achieve nearly 2.9x speedup of the dominant routines relative to the baseline. These exhibit almost linear strong scalability up to 64 threads, and thereafter some improvement with hyperthreading. At substantially fewer Watts, we achieve up to 1.7x speedup relative to the performance of 72 threads of a 36-core Haswell CPU and roughly equivalent performance to 112 threads of a 56-core Skylake scalable processor. These optimizations are expected to be of value for many other unstructured mesh PDE-based scientific applications as multi and many-core architecture evolves.

  6. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  7. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  8. Improving Battery Reactor Core Design Using Optimization Method

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2011-01-01

    The Battery Omnibus Reactor Integral System (BORIS) is a small modular fast reactor being designed at Seoul National University to satisfy various energy demands, to maintain inherent safety by liquid-metal coolant lead for natural circulation heat transport, and to improve power conversion efficiency with the Modular Optimal Balance Integral System (MOBIS) using the supercritical carbon dioxide as working fluid. This study is focused on developing the Neutronics Optimized Reactor Analysis (NORA) method that can quickly generate conceptual design of a battery reactor core by means of first principle calculations, which is part of the optimization process for reactor assembly design of BORIS

  9. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  10. Evaluating nuclear physics inputs in core-collapse supernova models

    Science.gov (United States)

    Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

    Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

  11. Apparatus for controlling nuclear core debris

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    Disclosed is an apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling

  12. Swarm intelligence of artificial bees applied to In-Core Fuel Management Optimization

    Energy Technology Data Exchange (ETDEWEB)

    Santos de Oliveira, Iona Maghali, E-mail: ioliveira@con.ufrj.br [Nuclear Engineering Program, Federal University of Rio de Janeiro, P.O. Box 68509, Zip Code 21945-970, Rio de Janeiro, RJ (Brazil); Schirru, Roberto, E-mail: schirru@lmp.ufrj.br [Nuclear Engineering Program, Federal University of Rio de Janeiro, P.O. Box 68509, Zip Code 21945-970, Rio de Janeiro, RJ (Brazil)

    2011-05-15

    Research highlights: > We present Artificial Bee Colony with Random Keys (ABCRK) for In-Core Fuel Management Optimization. > Its performance is examined through the optimization of a Brazilian '2-loop' PWR. > Feasibility of using ABCRK is shown against some well known population-based algorithms. > Additional advantage includes the utilization of fewer control parameters. - Abstract: Artificial Bee Colony (ABC) algorithm is a relatively new member of swarm intelligence. ABC tries to simulate the intelligent behavior of real honey bees in food foraging and can be used for solving continuous optimization and multi-dimensional numeric problems. This paper introduces the Artificial Bee Colony with Random Keys (ABCRK), a modified ABC algorithm for solving combinatorial problems such as the In-Core Fuel Management Optimization (ICFMO). The ICFMO is a hard combinatorial optimization problem in Nuclear Engineering which during many years has been solved by expert knowledge. It aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. As a consequence, the operation cost decreases and money is saved. In this study, ABCRK is used for optimizing the ICFMO problem of a Brazilian '2-loop' Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP) and the results obtained with the proposed algorithm are compared with those obtained by Genetic Algorithms (GA) and Particle Swarm Optimization (PSO). The results show that the performance of the ABCRK algorithm is better than or similar to that of other population-based algorithms, with the advantage of employing fewer control parameters.

  13. Swarm intelligence of artificial bees applied to In-Core Fuel Management Optimization

    International Nuclear Information System (INIS)

    Santos de Oliveira, Iona Maghali; Schirru, Roberto

    2011-01-01

    Research highlights: → We present Artificial Bee Colony with Random Keys (ABCRK) for In-Core Fuel Management Optimization. → Its performance is examined through the optimization of a Brazilian '2-loop' PWR. → Feasibility of using ABCRK is shown against some well known population-based algorithms. → Additional advantage includes the utilization of fewer control parameters. - Abstract: Artificial Bee Colony (ABC) algorithm is a relatively new member of swarm intelligence. ABC tries to simulate the intelligent behavior of real honey bees in food foraging and can be used for solving continuous optimization and multi-dimensional numeric problems. This paper introduces the Artificial Bee Colony with Random Keys (ABCRK), a modified ABC algorithm for solving combinatorial problems such as the In-Core Fuel Management Optimization (ICFMO). The ICFMO is a hard combinatorial optimization problem in Nuclear Engineering which during many years has been solved by expert knowledge. It aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. As a consequence, the operation cost decreases and money is saved. In this study, ABCRK is used for optimizing the ICFMO problem of a Brazilian '2-loop' Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP) and the results obtained with the proposed algorithm are compared with those obtained by Genetic Algorithms (GA) and Particle Swarm Optimization (PSO). The results show that the performance of the ABCRK algorithm is better than or similar to that of other population-based algorithms, with the advantage of employing fewer control parameters.

  14. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  15. Device for supporting a nuclear reactor core

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The core of a light-water reactor which is enclosed in a prestressed concrete pressure vessel and held within a diffuser basket is supported by a device consisting of a cylindrical shell which surrounds the basket and is rigidly fixed to a plurality of frusto-conical skirts having concurrent axes and located substantially at right angles to the axis of the reactor core. The small base of each skirt is rigidly fixed to the shell and the large base is anchored in openings formed in the reactor vessel for the penetration of coolant inlet and outlet pipes. The top portion of the shell is secured to the top portion of the diffuser basket, a flat surface being formed on the shell at the point of connection with each frusto-conical skirt so as to ensure rigid suspension while permitting thermal expansion

  16. A common and optimized age scale for Antarctic ice cores

    Science.gov (United States)

    Parrenin, F.; Veres, D.; Landais, A.; Bazin, L.; Lemieux-Dudon, B.; Toye Mahamadou Kele, H.; Wolff, E.; Martinerie, P.

    2012-04-01

    Dating ice cores is a complex problem because 1) there is a age shift between the gas bubbles and the surrounding ice 2) there are many different ice cores which can be synchronized with various proxies and 3) there are many methods to date the ice and the gas bubbles, each with advantages and drawbacks. These methods fall into the following categories: 1) Ice flow (for the ice) and firn densification modelling (for the gas bubbles); 2) Comparison of ice core proxies with insolation variations (so-called orbital tuning methods); 3) Comparison of ice core proxies with other well dated archives; 4) Identification of well-dated horizons, such as tephra layers or geomagnetic anomalies. Recently, an new dating tool has been developped (DATICE, Lemieux-Dudon et al., 2010), to take into account all the different dating information into account and produce a common and optimal chronology for ice cores with estimated confidence intervals. In this talk we will review the different dating information for Antarctic ice cores and show how the DATICE tool can be applied.

  17. Thermal barrier and support for nuclear reactor fuel core

    International Nuclear Information System (INIS)

    Betts, W.S. Jr.; Pickering, J.L.; Black, W.E.

    1987-01-01

    A nuclear reactor is described having a thermal barrier for supporting a fuel column of a nuclear reactor core within a reactor vessel having a fixed rigid metal liner. The fuel column has a refractory post extending downward. The thermal barrier comprises, in combination, a metallic core support having an interior chamber secured to the metal liner; fibrous thermal insulation material covering the metal liner and surrounding the metallic core support; means associated with the metallic core support and resting on the top for locating and supporting the full column post; and a column of ceramic material located within the interior chamber of the metallic core support, the height of the column is less than the height of the metallic core support so that the ceramic column will engage the means for locating and supporting the fuel column post only upon plastic deformation of the metallic core support; the core support comprises a metallic cylinder and the ceramic column comprises coaxially aligned ceramic pads. Each pad has a hole located within the metallic cylinder by means of a ceramic post passing through the holes in the pads

  18. Spring unit especially intended for a nuclear reactor core

    International Nuclear Information System (INIS)

    Brown, S.J.; Gorholt, Wilhelm.

    1977-01-01

    This invention relates to a spring unit or a group of springs bearing up a sprung mass against an unsprung mass. For instance, a gas cooled high temperature nuclear reactor includes a core of relatively complex structure supported inside a casing or vessel forming a shielded cavity enclosing the reactor core. This core can be assembled from a large number of graphite blocks of different sizes and shapes joined together to form a column. The blocks of each column can be fixed together so as to form together a loose side support. Under the effect of thermal expansion and contraction, shrinkage resulting from irradiation, the effects of pressure and the contraction and creep of the reactor vessel, it is not possible to confine all the columns of the reactor core in a cylindrical rigid structure. Further, the working of the nuclear reactor requires that the reactivity monitoring components may be inserted at any time in the reactor core. A standard process consists in mounting this loosely assembled reactor core in a floating manner by keeping it away from the vessel enclosure around it by means of a number of springs fitted between the lateral surfaces of the core unit and the reactor vessel. The core may be considered as a spring supported mass whereas, relatively, the reactor vessel is a mass that is not flexibly supported [fr

  19. Multi-objective and multi-physics optimization methodology for SFR core: application to CFV concept

    International Nuclear Information System (INIS)

    Fabbris, Olivier

    2014-01-01

    Nuclear reactor core design is a highly multidisciplinary task where neutronics, thermal-hydraulics, fuel thermo-mechanics and fuel cycle are involved. The problem is moreover multi-objective (several performances) and highly dimensional (several tens of design parameters).As the reference deterministic calculation codes for core characterization require important computing resources, the classical design method is not well suited to investigate and optimize new innovative core concepts. To cope with these difficulties, a new methodology has been developed in this thesis. Our work is based on the development and validation of simplified neutronics and thermal-hydraulics calculation schemes allowing the full characterization of Sodium-cooled Fast Reactor core regarding both neutronics performances and behavior during thermal hydraulic dimensioning transients.The developed methodology uses surrogate models (or meta-models) able to replace the neutronics and thermal-hydraulics calculation chain. Advanced mathematical methods for the design of experiment, building and validation of meta-models allows substituting this calculation chain by regression models with high prediction capabilities.The methodology is applied on a very large design space to a challenging core called CFV (French acronym for low void effect core) with a large gain on the sodium void effect. Global sensitivity analysis leads to identify the significant design parameters on the core design and its behavior during unprotected transient which can lead to severe accidents. Multi-objective optimizations lead to alternative core configurations with significantly improved performances. Validation results demonstrate the relevance of the methodology at the pre-design stage of a Sodium-cooled Fast Reactor core. (author) [fr

  20. The online simulation of core physics in nuclear power plant

    International Nuclear Information System (INIS)

    Zhao Qiang

    2005-01-01

    The three-dimensional power distribution in core is one of the most important status variables of nuclear reactor. In order to monitor the 3-D in core power distribution timely and accurately, the online simulation system of core physics was designed in the paper. This system combines core physics simulation with the data, which is from the plant and reactor instrumentation. The design of the system consists of the hardware part and the software part. The online simulation system consists of a main simulation computer and a simulation operation station. The online simulation system software includes of the real-time simulation support software, the system communication software, the simulation program and the simulation interface software. Two-group and three-dimensional neutron kinetics model with six groups delayed neutrons was used in the real-time simulation of nuclear reactor core physics. According to the characteristics of the nuclear reactor, the core was divided into many nodes. Resolving the neutron equation, the method of separate variables was used. The input data from the plant and reactor instrumentation system consist of core thermal power, loop temperatures and pressure, control rod positions, boron concentration, core exit thermocouple data, Excore detector signals, in core flux detectors signals. There are two purposes using the data, one is to ensure that the model is as close as the current actual reactor condition, and the other is to calibrate the calculated power distribution. In this paper, the scheme of the online simulation system was introduced. Under the real-time simulation support system, the simulation program is being compiled. Compared with the actual operational data, the elementary simulation results were reasonable and correct. (author)

  1. Design and axial optimization of nuclear fuel for BWR reactors

    International Nuclear Information System (INIS)

    Garcia V, M.A.

    2006-01-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  2. Fuel optimization of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Liao Zejun; Li Zhuoqun; Kong Deping; Xue Xincai; Wang Shiwei

    2010-01-01

    Based on the design practice of the fuel replacement of Qin Shan nuclear power plant, this document effectively analyzes the shortcomings of current replacement design of Qin Shan. To address these shortcomings, this document successfully implements the 300 MW fuel optimization program from fuel replacement. fuel improvement and experimentation ,and achieves great economic results. (authors)

  3. Engineering application of in-core fuel management optimization code with CSA algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zhihong; Hu, Yongming [INET, Tsinghua university, Beijing 100084 (China)

    2009-06-15

    PWR in-core loading (reloading) pattern optimization is a complex combined problem. An excellent fuel management optimization code can greatly improve the efficiency of core reloading design, and bring economic and safety benefits. Today many optimization codes with experiences or searching algorithms (such as SA, GA, ANN, ACO) have been developed, while how to improve their searching efficiency and engineering usability still needs further research. CSA (Characteristic Statistic Algorithm) is a global optimization algorithm with high efficiency developed by our team. The performance of CSA has been proved on many problems (such as Traveling Salesman Problems). The idea of CSA is to induce searching direction by the statistic distribution of characteristic values. This algorithm is quite suitable for fuel management optimization. Optimization code with CSA has been developed and was used on many core models. The research in this paper is to improve the engineering usability of CSA code according to all the actual engineering requirements. Many new improvements have been completed in this code, such as: 1. Considering the asymmetry of burn-up in one assembly, the rotation of each assembly is considered as new optimization variables in this code. 2. Worth of control rods must satisfy the given constraint, so some relative modifications are added into optimization code. 3. To deal with the combination of alternate cycles, multi-cycle optimization is considered in this code. 4. To confirm the accuracy of optimization results, many identifications of the physics calculation module in this code have been done, and the parameters of optimization schemes are checked by SCIENCE code. The improved optimization code with CSA has been used on Qinshan nuclear plant of China. The reloading of cycle 7, 8, 9 (12 months, no burnable poisons) and the 18 months equilibrium cycle (with burnable poisons) reloading are optimized. At last, many optimized schemes are found by CSA code

  4. Perturbation theory in nuclear fuel management optimization

    International Nuclear Information System (INIS)

    Ho, L.W.; Rohach, A.F.

    1982-01-01

    Perturbation theory along with a binary fuel shuffling technique is applied to predict the effects of various core configurations and, hence, the optimization of in-core fuel management. The computer code FULMNT has been developed to shuffle the fuel assemblies in search of the lowest possible power peaking factor. An iteration approach is used in the search routine. A two-group diffusion theory method is used to obtain the power distribution for the iterations. A comparison of the results of this method with other methods shows that this approach can save computer time and obtain better power peaking factors. The code also has a burnup capability that can be used to check power peaking throughout the core life

  5. Temperature measurements inside nuclear reactor cores

    International Nuclear Information System (INIS)

    Tarassenko, Serge

    1969-11-01

    Non negligible errors may happen in nuclear reactor temperature measurements using magnesium oxide insulated and stainless steel sheathed micro-wire thermocouples, when these thermometric lines are placed under operational conditions typical of electrical power stations. The present work shows that this error is principally due to electrical hysteresis and polarization phenomena in the insulator subjected to the strong fields generated by common-mode voltages. These phenomena favour the unsymmetrical common-mode current flow and thus lead to the differential-mode voltage generation which is superposing on the thermoelectric hot junction potential. A calculation and an experimental approach make possible the importance of the magnesium oxide insulating characteristics, the hot junction insulation, the choice of the main circuits in the data processing equipment as well as the galvanic isolation performances and the common-mode rejection features of all the measurement circuits. A justification is thereby given for the severe conditions imposed for the acceptance of thermoelectric materials; some particular precautions to be taken are described, as well as the high performance characteristics which have to be taken into account in choosing measurement systems linked to thermometric circuits with sheathed micro-wire thermocouples. (author) [fr

  6. Status of core nuclear design technology for future fuel

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Jung, Hyung Guk; Noh, Jae Man; Kim, Yeong Il; Kim, Taek Kyum; Gil, Choong Sup; Kim, Jung Do; Kim, Young Jin; Sohn, Dong Seong

    1997-01-01

    The effective utilization of nuclear resource is more important factor to be considered in the design of next generation PWR in addition to the epochal consideration on economics and safety. Assuming that MOX fuel can be considered as one of the future fuel corresponding to the above request, the establishment of basic technology for the MOX core design has been performed : : the specification of the technical problem through the preliminary core design and nuclear characteristic analysis of MOX, the development and verification of the neutron library for lattice code, and the acquisition of data to be used for verification of lattice and core analysis codes. The following further studies will be done in future: detailed verification of library E63LIB/A, development of the spectral history effect treatment module, extension of decay chain, development of new homogenization for the MOX fuel assembly. (author). 6 refs., 7 tabs., 2 figs

  7. Thermal radiation in gas core nuclear reactors for space propulsion

    International Nuclear Information System (INIS)

    Slutz, S.A.; Gauntt, R.O.; Harms, G.A.; Latham, T.; Roman, W.; Rodgers, R.J.

    1994-01-01

    A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs

  8. A multi-cycle optimization approach for low leakage in-core fuel management

    International Nuclear Information System (INIS)

    Cheng Pingdong; Shen Wei

    1999-01-01

    A new approach was developed to optimize pressurized waster reactor (PWR) low-leakage multi-cycle reload core design. The multi-cycle optimization process is carried out by the following three steps: The first step is a linear programming in search for an optimum power sharing distribution and optimum cycle length distribution for the successive several cycles to yield maximum multi-cycle total cycle length. In the second step, the fuel arrangement and burnable poison (BP) assignment are decoupled by using Haling power distribution and the optimum fuel arrangement is determined at the EOL in the absence of all BPs by employing a linear programming method or direct search method with objective function to force the calculated cycle length to be as close as possible to the optimum single cycle length obtained in the first step and with optimum power sharing distribution as additional constraints during optimization. In the third step, the BP assignment is optimized by the Flexible Tolerance Method (FTM) or linear programming method using the number of BP rods as control variable. The technology employed in the second and third steps was the usual decoupling method used in low-leakage core design. The first step was developed specially for multi-cycle optimization design and discussed in detail. Based on the proposed method a computer code MCYCO was encoded and tested for Qinshan Nuclear Power Plant (QNPP) low leakage reload core design. The multi-cycle optimization method developed, together with the program MCYCO, provides an applicable tool for solving the PWR low leakage reload core design problem

  9. A higher order depletion perturbation theory with application to in-core fuel management optimization

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1990-01-01

    Perturbation techniques utilized in reactor analysis have recently been applied in the solution of the in-core nuclear fuel management optimization problem. The use of such methods is motivated by the need to evaluate many times over, the core physics characteristics of loading pattern solutions obtained through an optimization process, which is typically iterative. Perturbation theory provides an efficient alternative to the prohibitively expensive, repetitive solutions of the system few-group neutron diffusion equations required in solving the fuel placement problem. A primary concern in the use of such methods is the control of perturbation errors arising during the fuel shuffling process. First-order accurate models inevitably resort to undue restriction of fuel movement during the optimization process to control these errors. Higher order perturbation theory models have the potential to overcome such limitations, which may result in the identification of local versus global optima. An accurate, computationally efficient reactor physics model based on higher order perturbation theory and geared toward the needs of large-scale in-core fuel management optimization is presented in this paper

  10. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  11. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  12. Sensitivity of control times in function of core parameters and oscillations control in thermal nuclear systems

    International Nuclear Information System (INIS)

    Amorim, E.S. do; D'Oliveira, A.B.; Galvao, O.B.; Oyama, K.

    1981-03-01

    Sensitivity of control times to variation of a thermal reactor core parameters is defined by suitable changes in the power coefficient, core size and fuel enrichment. A control strategy is developed based on control theory concepts and on considerations of the physics of the problem. Digital diffusion theory simulation is described which tends to verify the control concepts considered, face dumped oscillations introduced in one thermal nuclear power system. The effectivity of the control actions, in terms of eliminating oscillations, provided guidelines for the working-group engaged in the analysis of the control rods and its optimal performance. (Author) [pt

  13. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  14. Nuclear Human Resources Development Program using Educational Core Simulator

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Hong, Soon Kwan

    2015-01-01

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides

  15. Nuclear Human Resources Development Program using Educational Core Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yu Sun; Hong, Soon Kwan [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    KHNP-CRI(Korea Hydro and Nuclear Power Co.-Central Research Institute) has redesigned the existing Core Simulator(CoSi) used as a sort of training tools for reactor engineers in operating nuclear power plant to support Nuclear Human Resources Development (NHRD) Program focusing on the nuclear department of Dalat university in Vietnam. This program has been supported by MOTIE in Korea and cooperated with KNA(Korea Nuclear Association for International Cooperation) and HYU(Hanyang University) for enhancing the nuclear human resources of potential country in consideration with Korean Nuclear Power Plant as a next candidate energy sources. KHNP-CRI has provided Edu-CoSi to Dalat University in Vietnam in order to support Nuclear Human Resources Development Program in Vietnam. Job Qualification Certificates Program in KHNP is utilized to design a training course for Vietnamese faculty and student of Dalat University. Successfully, knowhow on lecturing the ZPPT performance, training and maintaining Edu-CoSi hardware are transferred by several training courses which KHNP-CRI provides.

  16. Enhanced core monitoring system for Browns Ferry Nuclear Plant

    International Nuclear Information System (INIS)

    Lindsey, R.S.

    1980-01-01

    A system of computer hardware and software is being developed to supplement the process computers at Browns Ferry Nuclear Plant in the area of reactor core monitoring. All data stored in the process computers will be made available through a data link to an onsite minicomputer which will store and edit the data for engineering and operations personnel. Important core parameters will be effectively displayed on color graphic CRT terminals using techniques such as blinking, shading, and color coding to emphasize significant values. This data will also be made available to Tennessee Valley Authority's Chattanooga central office support groups through a data network between existing computers

  17. Performance of Estimation of distribution algorithm for initial core loading optimization of AHWR-LEU

    International Nuclear Information System (INIS)

    Thakur, Amit; Singh, Baltej; Gupta, Anurag; Duggal, Vibhuti; Bhatt, Kislay; Krishnani, P.D.

    2016-01-01

    Highlights: • EDA has been applied to optimize initial core of AHWR-LEU. • Suitable value of weighing factor ‘α’ and population size in EDA was estimated. • The effect of varying initial distribution function on optimized solution was studied. • For comparison, Genetic algorithm was also applied. - Abstract: Population based evolutionary algorithms now form an integral part of fuel management in nuclear reactors and are frequently being used for fuel loading pattern optimization (LPO) problems. In this paper we have applied Estimation of distribution algorithm (EDA) to optimize initial core loading pattern (LP) of AHWR-LEU. In EDA, new solutions are generated by sampling the probability distribution model estimated from the selected best candidate solutions. The weighing factor ‘α’ decides the fraction of current best solution for updating the probability distribution function after each generation. A wider use of EDA warrants a comprehensive study on parameters like population size, weighing factor ‘α’ and initial probability distribution function. In the present study, we have done an extensive analysis on these parameters (population size, weighing factor ‘α’ and initial probability distribution function) in EDA. It is observed that choosing a very small value of ‘α’ may limit the search of optimized solutions in the near vicinity of initial probability distribution function and better loading patterns which are away from initial distribution function may not be considered with due weightage. It is also observed that increasing the population size improves the optimized loading pattern, however the algorithm still fails if the initial distribution function is not close to the expected optimized solution. We have tried to find out the suitable values for ‘α’ and population size to be considered for AHWR-LEU initial core loading pattern optimization problem. For sake of comparison and completeness, we have also addressed the

  18. MTR core loading pattern optimization using burnup dependent group constants

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

    Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.

  19. Quantification of cost of margin associated with in-core nuclear fuel management for a PWR

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1989-01-01

    The problem of in-core nuclear fuel management optimization is discussed. The problem is to determine the location of core material, such as the fuel and burnable poisons, so as to minimize (maximize) a stated objective within engineering constraints. Typical objectives include maximization of cycle energy production or discharged fuel exposure, and minimization of power peaking factor or reactor vessel fluence. Constraints include discharge burnup limits and one or more of the possible objectives if not selected as the objective. The optimization problem can be characterized as a large combinatorial problem with nonlinear objective function and constraints, which are likely to be active. The authors have elected to employ the integer Monte Carlo programming method to address this optimization problem because of the just-noted problem characteristics. To evaluate the core physics characteristics as a function of fuel loading pattern, second-order accurate perturbation theory is employed with successive application to improve estimates of the optimum loading pattern. No constraints on fuel movement other than requiring quarter-core symmetry were imposed. In this paper the authors employed this methodology to address a related problem. The problem being addressed can be stated as What is the cost associated with margin? Specifically, they wish to assign some financial value in terms of increased levelized fuel cycle cost associated with an increase in core margin of some type, such as power peaking factor

  20. Self-adaptive global best harmony search algorithm applied to reactor core fuel management optimization

    International Nuclear Information System (INIS)

    Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.; Valavi, K.

    2013-01-01

    Highlights: • SGHS enhanced the convergence rate of LPO using some improvements in comparison to basic HS and GHS. • SGHS optimization algorithm obtained averagely better fitness relative to basic HS and GHS algorithms. • Upshot of the SGHS implementation in the LPO reveals its flexibility, efficiency and reliability. - Abstract: The aim of this work is to apply the new developed optimization algorithm, Self-adaptive Global best Harmony Search (SGHS), for PWRs fuel management optimization. SGHS algorithm has some modifications in comparison with basic Harmony Search (HS) and Global-best Harmony Search (GHS) algorithms such as dynamically change of parameters. For the demonstration of SGHS ability to find an optimal configuration of fuel assemblies, basic Harmony Search (HS) and Global-best Harmony Search (GHS) algorithms also have been developed and investigated. For this purpose, Self-adaptive Global best Harmony Search Nodal Expansion package (SGHSNE) has been developed implementing HS, GHS and SGHS optimization algorithms for the fuel management operation of nuclear reactor cores. This package uses developed average current nodal expansion code which solves the multi group diffusion equation by employment of first and second orders of Nodal Expansion Method (NEM) for two dimensional, hexagonal and rectangular geometries, respectively, by one node per a FA. Loading pattern optimization was performed using SGHSNE package for some test cases to present the SGHS algorithm capability in converging to near optimal loading pattern. Results indicate that the convergence rate and reliability of the SGHS method are quite promising and practically, SGHS improves the quality of loading pattern optimization results relative to HS and GHS algorithms. As a result, it has the potential to be used in the other nuclear engineering optimization problems

  1. An Interval Bound Algorithm of optimizing reactor core loading pattern by using reactivity interval schema

    International Nuclear Information System (INIS)

    Gong Zhaohu; Wang Kan; Yao Dong

    2011-01-01

    Highlights: → We present a new Loading Pattern Optimization method - Interval Bound Algorithm (IBA). → IBA directly uses the reactivity of fuel assemblies and burnable poison. → IBA can optimize fuel assembly orientation in a coupled way. → Numerical experiment shows that IBA outperforms genetic algorithm and engineers. → We devise DDWF technique to deal with multiple objectives and constraints. - Abstract: In order to optimize the core loading pattern in Nuclear Power Plants, the paper presents a new optimization method - Interval Bound Algorithm (IBA). Similar to the typical population based algorithms, e.g. genetic algorithm, IBA maintains a population of solutions and evolves them during the optimization process. IBA acquires the solution by statistical learning and sampling the control variable intervals of the population in each iteration. The control variables are the transforms of the reactivity of fuel assemblies or the worth of burnable poisons, which are the crucial heuristic information for loading pattern optimization problems. IBA can deal with the relationship between the dependent variables by defining the control variables. Based on the IBA algorithm, a parallel Loading Pattern Optimization code, named IBALPO, has been developed. To deal with multiple objectives and constraints, the Dynamic Discontinuous Weight Factors (DDWF) for the fitness function have been used in IBALPO. Finally, the code system has been used to solve a realistic reloading problem and a better pattern has been obtained compared with the ones searched by engineers and genetic algorithm, thus the performance of the code is proved.

  2. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  3. Real-time advanced nuclear reactor core model

    International Nuclear Information System (INIS)

    Koclas, J.; Friedman, F.; Paquette, C.; Vivier, P.

    1990-01-01

    The paper describes a multi-nodal advanced nuclear reactor core model. The model is based on application of modern equivalence theory to the solution of neutron diffusion equation in real time employing the finite differences method. The use of equivalence theory allows the application of the finite differences method to cores divided into hundreds of nodes, as opposed to the much finer divisions (in the order of ten thousands of nodes) where the unmodified method is currently applied. As a result the model can be used for modelling of the core kinetics for real time full scope training simulators. Results of benchmarks, validate the basic assumptions of the model and its applicability to real-time simulation. (orig./HP)

  4. Development of Core Monitoring System for Nuclear Power Plants (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

  5. In-core fuel management for nuclear reactor

    International Nuclear Information System (INIS)

    Ross, M.F.; Visner, S.

    1986-01-01

    This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

  6. Westinghouse Nuclear Core Design Training Center - a design simulator

    International Nuclear Information System (INIS)

    Altomare, S.; Pritchett, J.; Altman, D.

    1992-01-01

    The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

  7. Nuclear physics: the core of matter, the fuel of stars

    International Nuclear Information System (INIS)

    Schiffer, J.P.

    1999-01-01

    Dramatic progress has been made in all branches of physics since the National Research Council's 1986 decadal survey of the field. The Physics in a New Era series explores these advances and looks ahead to future goals. The series includes assessments of the major subfields and reports on several smaller subfields, and preparation has begun on an overview volume on the unity of physics, its relationships to other fields, and its contributions to national needs. Nuclear Physics is the latest volume of the series. The book describes current activity in understanding nuclear structure and symmetries, the behavior of matter at extreme densities, the role of nuclear physics in astrophysics and cosmology, and the instrumentation and facilities used by the field. It makes recommendations on the resources needed for experimental and theoretical advances in the coming decade. Nuclear physics addresses the nature of matter making up 99.9 percent of the mass of our everyday world. It explores the nuclear reactions that fuel the stars, including our Sun, which provides the energy for all life on Earth. The field of nuclear physics encompasses some 3,000 experimental and theoretical researchers who work at universities and national laboratories across the United States, as well as the experimental facilities and infrastructure that allow these researchers to address the outstanding scientific questions facing us. This report provides an overview of the frontiers of nuclear physics as we enter the next millennium, with special attention to the state of the science in the United States.The current frontiers of nuclear physics involve fundamental and rapidly evolving issues. One is understanding the structure and behavior of strongly interacting matter in terms of its basic constituents, quarks and gluons, over a wide range of conditions - from normal nuclear matter to the dense cores of neutron stars, and to the Big Bang that was the birth of the universe. Another is to describe

  8. Stochastic Optimization for Nuclear Facility Deployment Scenarios

    Science.gov (United States)

    Hays, Ross Daniel

    Single-use, low-enriched uranium oxide fuel, consumed through several cycles in a light-water reactor (LWR) before being disposed, has become the dominant source of commercial-scale nuclear electric generation in the United States and throughout the world. However, it is not without its drawbacks and is not the only potential nuclear fuel cycle available. Numerous alternative fuel cycles have been proposed at various times which, through the use of different reactor and recycling technologies, offer to counteract many of the perceived shortcomings with regards to waste management, resource utilization, and proliferation resistance. However, due to the varying maturity levels of these technologies, the complicated material flow feedback interactions their use would require, and the large capital investments in the current technology, one should not deploy these advanced designs without first investigating the potential costs and benefits of so doing. As the interactions among these systems can be complicated, and the ways in which they may be deployed are many, the application of automated numerical optimization to the simulation of the fuel cycle could potentially be of great benefit to researchers and interested policy planners. To investigate the potential of these methods, a computational program has been developed that applies a parallel, multi-objective simulated annealing algorithm to a computational optimization problem defined by a library of relevant objective functions applied to the Ver ifiable Fuel Cycle Simulati on Model (VISION, developed at the Idaho National Laboratory). The VISION model, when given a specified fuel cycle deployment scenario, computes the numbers and types of, and construction, operation, and utilization schedules for, the nuclear facilities required to meet a predetermined electric power demand function. Additionally, it calculates the location and composition of the nuclear fuels within the fuel cycle, from initial mining through

  9. 3D Core Model for simulation of nuclear power plants: Simulation requirements, model features, and validation

    International Nuclear Information System (INIS)

    Zerbino, H.

    1999-01-01

    In 1994-1996, Thomson Training and Simulation (TT and S) earned out the D50 Project, which involved the design and construction of optimized replica simulators for one Dutch and three German Nuclear Power Plants. It was recognized early on that the faithful reproduction of the Siemens reactor control and protection systems would impose extremely stringent demands on the simulation models, particularly the Core physics and the RCS thermohydraulics. The quality of the models, and their thorough validation, were thus essential. The present paper describes the main features of the fully 3D Core model implemented by TT and S, and its extensive validation campaign, which was defined in extremely positive collaboration with the Customer and the Core Data suppliers. (author)

  10. Device for measuring flow rate in a nuclear reactor core

    International Nuclear Information System (INIS)

    Hamano, Jiro.

    1980-01-01

    Purpose: To always calculate core flow rate automatically and accurately in BWR type nuclear power plants. Constitution: Jet pumps are provided to the recycling pump and to the inside of the pressure vessel of a nuclear reactor. The jet pumps comprise a plurality of calibrated jet pumps for forcively convecting the coolants and a plurality of not calibrated jet pumps in order to cool the heat generated in the reactor core. The difference in the pressures between the upper and the lower portions in both of the jet pumps is measured by difference pressure transducers. Further, a thermo-sensitive element is provided to measure the temperature of recycling water at the inlet of the recycling pump. The output signal from the difference pressure transducer is inputted to a process computer, calculated periodically based on predetermined calculation equations, compensated for the temperature by a recycling water temperature signal and outputted as a core flow rate signal to a recoder. The signal is also used for the power distribution calculation in the process computer and the minimum limit power ratio as the thermal limit value for the fuels is outputted. (Furukawa, Y.)

  11. In-Core Fuel Management with Biased Multiobjective Function Optimization

    International Nuclear Information System (INIS)

    Shatilla, Youssef A.; Little, David C.; Penkrot, Jack A.; Holland, Richard Andrew

    2000-01-01

    The capability of biased multiobjective function optimization has been added to the Westinghouse Electric Company's (Westinghouse's) Advanced Loading Pattern Search code (ALPS). The search process, given a user-defined set of design constraints, proceeds to minimize a global parameter called the total value associated with constraints compliance (VACC), an importance-weighted measure of the deviation from limit and/or margin target. The search process takes into consideration two equally important user-defined factors while minimizing the VACC, namely, the relative importance of each constraint with respect to the others and the optimization of each constraint according to its own objective function. Hence, trading off margin-to-design limits from where it is abundantly available to where it is badly needed can now be accomplished. Two practical methods are provided to the user for input of constraints and associated objective functions. One consists of establishing design limits based on traditional core design parameters such as assembly/pin burnup, power, or reactivity. The second method allows the user to write a program, or script, to define a logic not possible through ordinary means. This method of script writing was made possible through the application resident compiler feature of the technical user language integration processor (tulip), developed at Westinghouse. For the optimization problems studied, ALPS not only produced candidate loading patterns (LPs) that met all of the conflicting design constraints, but in cases where the design appeared to be over constrained gave a wide range of LPs that came very close to meeting all the constraints based on the associated objective functions

  12. Nuclear piston engine and pulsed gaseous core reactor power systems

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

    The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

  13. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-12-01

    Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

  14. Review of coaxial flow gas core nuclear rocket fluid mechanics

    International Nuclear Information System (INIS)

    Weinstein, H.

    1976-01-01

    In a prematurely aborted attempt to demonstrate the feasibility of using a gas core nuclear reactor as a rocket engine, NASA initiated a number of studies on the relevant fluid mechanics problems. These studies were carried out at NASA laboratories, universities and industrial research laboratories. Because of the relatively sudden termination of most of this work, a unified overview was never presented which demonstrated the accomplishments of the program and pointed out the areas where additional work was required for a full understanding of the cavity flow. This review attempts to fulfill a part of this need in two important areas

  15. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  16. Automated software analysis of nuclear core discharge data

    International Nuclear Information System (INIS)

    Larson, T.W.; Halbig, J.K.; Howell, J.A.; Eccleston, G.W.; Klosterbuer, S.F.

    1993-03-01

    Monitoring the fueling process of an on-load nuclear reactor is a full-time job for nuclear safeguarding agencies. Nuclear core discharge monitors (CDMS) can provide continuous, unattended recording of the reactor's fueling activity for later, qualitative review by a safeguards inspector. A quantitative analysis of this collected data could prove to be a great asset to inspectors because more information can be extracted from the data and the analysis time can be reduced considerably. This paper presents a prototype for an automated software analysis system capable of identifying when fuel bundle pushes occurred and monitoring the power level of the reactor. Neural network models were developed for calculating the region on the reactor face from which the fuel was discharged and predicting the burnup. These models were created and tested using actual data collected from a CDM system at an on-load reactor facility. Collectively, these automated quantitative analysis programs could help safeguarding agencies to gain a better perspective on the complete picture of the fueling activity of an on-load nuclear reactor. This type of system can provide a cost-effective solution for automated monitoring of on-load reactors significantly reducing time and effort

  17. Development of conceptual nuclear design of 10MWt research reactor core

    International Nuclear Information System (INIS)

    Kim, M. H.; Lim, J. Y.; Win, Naing; Park, J. M.

    2008-03-01

    KAERI has been devoted to develop export-oriented research reactors for a growing world-wide demand of new research reactor construction. Their ambition is that design of Korean research reactor must be competitive in commercial and technological based on the experience of the HANARO core design concept with thermal power of 30MW. They are developing a new research reactor named Advanced HANARO research Reactor (AHR) with thermal power of 20 MW. KAERI has export records of nuclear technology. In 1954-1967 two series of pool type research reactors based on the Russian design, VVR type and IRT type, have been constructed and commissioned in some countries as well as Russia. Nowadays Russian design is introducing again for export to developing countries such as Union of Myanmar. Therefore the objective of this research is that to build and innovative 10 MW research reactor core design based on the concept of HANARO core design to be competitive with Russian research reactor core design. system tool of HELIOS was used at the first stage in both cases which are research reactor using tubular type fuel assemblies and that reactor using pin type fuel assemblies. The reference core design of first kind of research reactor includes one in-core irradiation site at the core center. The neutron flux evaluations for core as well as reflector region were done through logical consistency of neutron flux distributions for individual assemblies. In order to find the optimum design, the parametric studies were carried out for assembly pitch, active fuel length, number of fuel ring in each assembly and so on. Design result shows the feasibility to have high neutron flux at in-core irradiation site. The second kind of research reactor is used the same kind of assemblies as HANARO and hence there is no optimization about basic design parameters. That core has only difference composition of assemblies and smaller specific power than HANARO. Since it is a reference core at first stage

  18. Pipe support optimization in nuclear power plants

    International Nuclear Information System (INIS)

    Cleveland, A.B.; Kalyanam, N.

    1984-01-01

    A typical 1000 MWe nuclear power plant consists of 80,000 to 100,000 feet of piping which must be designed to withstand earthquake shock. For the required ground motion, seismic response spectra are developed for safety-related structures. These curves are used in the dynamic analysis of piping systems with pipe-stress analysis computer codes. To satisfy applicable Code requirements, the piping systems also require analysis for weight, thermal and possibly other lasting conditions. Bechtel Power Corporation has developed a design program called SLAM (Support Location Algorithm) for optimizing pipe support locations and types (rigid, spring, snubber, axial, lateral, etc.) while satisfying userspecified parameters such as locations, load combinations, stress and load allowables, pipe displacement and cost. This paper describes SLAM, its features, applications and benefits

  19. Global physical and numerical stability of a nuclear reactor core

    International Nuclear Information System (INIS)

    Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

    2005-01-01

    Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

  20. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  1. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

  2. Optimization of Nuclear Reactor power Distribution using Genetic Algorithm

    International Nuclear Information System (INIS)

    Kim, Hyu Chan

    1996-02-01

    The main purpose of study is to develop a computer code named as 'MGA-SCOUPE' which can determine an optimal fuel-loading pattern for the nuclear reactor. The developed code, MGA-SCOUPE, automatically lots of searches for the globally optimum solutions based upon the modified Genetic Algorithm(MGA). The optimization goal of the MGA-SCOUPE is (1) the minimization of the deviations in the power peaking factors both at BOC and EOC, and (2) the maximization of the average burnup ration at EOC of the total fuel assemblies. For the reactor core calculation module in the MGA-SCOUPE, the SCOUPE code was partially modified and used. It had been developed originally in MIT and has been used currently in Kyung Hee University. The application of the MGA-SCOUPE to KORI 4-4 Cycle Model show several satisfactory results. Among them, two dominant improvements compared with the SCOUPE code can be summarized as follow: - The MGA-SCOUPE removes the user-dependency problem of the SCOUPE in the optimal loading pattern searches. Therefore, the searching process in the MGA-SCOUPE can be easily automated. - The final fuel loading pattern obtained by the MGA-SCOUPE shows 25.8%, 18.7% reduced standard deviations of the power peaking factors both at BOC and EOC, and 45% increased avg. burnup ratio at EOC compare with those of the SCOUPE

  3. Parallel particle swarm optimization algorithm in nuclear problems

    International Nuclear Information System (INIS)

    Waintraub, Marcel; Pereira, Claudio M.N.A.; Schirru, Roberto

    2009-01-01

    Particle Swarm Optimization (PSO) is a population-based metaheuristic (PBM), in which solution candidates evolve through simulation of a simplified social adaptation model. Putting together robustness, efficiency and simplicity, PSO has gained great popularity. Many successful applications of PSO are reported, in which PSO demonstrated to have advantages over other well-established PBM. However, computational costs are still a great constraint for PSO, as well as for all other PBMs, especially in optimization problems with time consuming objective functions. To overcome such difficulty, parallel computation has been used. The default advantage of parallel PSO (PPSO) is the reduction of computational time. Master-slave approaches, exploring this characteristic are the most investigated. However, much more should be expected. It is known that PSO may be improved by more elaborated neighborhood topologies. Hence, in this work, we develop several different PPSO algorithms exploring the advantages of enhanced neighborhood topologies implemented by communication strategies in multiprocessor architectures. The proposed PPSOs have been applied to two complex and time consuming nuclear engineering problems: reactor core design and fuel reload optimization. After exhaustive experiments, it has been concluded that: PPSO still improves solutions after many thousands of iterations, making prohibitive the efficient use of serial (non-parallel) PSO in such kind of realworld problems; and PPSO with more elaborated communication strategies demonstrated to be more efficient and robust than the master-slave model. Advantages and peculiarities of each model are carefully discussed in this work. (author)

  4. A simulator-independent optimization tool based on genetic algorithm applied to nuclear reactor design

    International Nuclear Information System (INIS)

    Abreu Pereira, Claudio Marcio Nascimento do; Schirru, Roberto; Martinez, Aquilino Senra

    1999-01-01

    Here is presented an engineering optimization tool based on a genetic algorithm, implemented according to the method proposed in recent work that has demonstrated the feasibility of the use of this technique in nuclear reactor core designs. The tool is simulator-independent in the sense that it can be customized to use most of the simulators which have the input parameters read from formatted text files and the outputs also written from a text file. As the nuclear reactor simulators generally use such kind of interface, the proposed tool plays an important role in nuclear reactor designs. Research reactors may often use non-conventional design approaches, causing different situations that may lead the nuclear engineer to face new optimization problems. In this case, a good optimization technique, together with its customizing facility and a friendly man-machine interface could be very interesting. Here, the tool is described and some advantages are outlined. (author)

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

    International Nuclear Information System (INIS)

    Alvarenga de Moura Meneses, Anderson; Rancoita, Paola; Schirru, Roberto; Gambardella, Luca Maria

    2010-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alvarenga de Moura Meneses, Anderson, E-mail: ameneses@lmp.ufrj.b [Federal University of Rio de Janeiro, COPPE, Nuclear Engineering Program, CP 68509, CEP 21.941-972, Rio de Janeiro, RJ (Brazil); Rancoita, Paola [IDSIA (Dalle Molle Institute for Artificial Intelligence), Galleria 2, 6982 Manno-Lugano, TI (Switzerland); Mathematics Department, Universita degli Studi di Milano (Italy); Schirru, Roberto [Federal University of Rio de Janeiro, COPPE, Nuclear Engineering Program, CP 68509, CEP 21.941-972, Rio de Janeiro, RJ (Brazil); Gambardella, Luca Maria [IDSIA (Dalle Molle Institute for Artificial Intelligence), Galleria 2, 6982 Manno-Lugano, TI (Switzerland)

    2010-11-15

    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.

  7. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes.

  8. Regulatory Audit Activities on Nuclear Design of Reactor Cores

    International Nuclear Information System (INIS)

    Yang, Chae-Yong; Lee, Gil Soo; Lee, Jaejun; Kim, Gwan-Young; Bae, Moo-Hun

    2016-01-01

    Regulatory audit analyses are initiated on the purpose of deep knowledge, solving safety issues, being applied in the review of licensee's results. The current most important safety issue on nuclear design is to verify bias and uncertainty on reactor physics codes to examine the behaviors of high burnup fuel during rod ejection accident (REA) and LOCA, and now regulatory audits are concentrated on solving this issue. KINS develops regulatory audit tools on its own, and accepts ones verified from foreign countries. The independent audit tools are sometimes standardized through participating the international programs. New safety issues on nuclear design, reactor physics tests, advanced reactor core design are steadily raised, which are mainly drawn from the independent examination tools. It is some facing subjects for the regulators to find out the unidentified uncertainties in high burnup fuels and to systematically solve them. The safety margin on nuclear design might be clarified by precisely having independent tools and doing audit calculations by using them. SCALE-PARCS/COREDAX and the coupling with T-H code or fuel performance code would be certainly necessary for achieving these purposes

  9. Genetic algorithm with fuzzy clustering for optimization of nuclear reactor problems

    International Nuclear Information System (INIS)

    Machado, Marcelo Dornellas; Sacco, Wagner Figueiredo; Schirru, Roberto

    2000-01-01

    Genetic Algorithms (GAs) are biologically motivated adaptive systems which have been used, with good results, in function optimization. However, traditional GAs rapidly push an artificial population toward convergence. That is, all individuals in the population soon become nearly identical. Niching Methods allow genetic algorithms to maintain a population of diverse individuals. GAs that incorporate these methods are capable of locating multiple, optimal solutions within a single population. The purpose of this study is to introduce a new niching technique based on the fuzzy clustering method FCM, bearing in mind its eventual application in nuclear reactor related problems, specially the nuclear reactor core reload one, which has multiple solutions. tests are performed using widely known test functions and their results show that the new method is quite promising, specially to a future application in real world problems like the nuclear reactor core reload. (author)

  10. Utilization of niching methods of genetic algorithms in nuclear reactor problems optimization

    International Nuclear Information System (INIS)

    Sacco, Wagner Figueiredo; Schirru, Roberto

    2000-01-01

    Genetic Algorithms (GAs) are biologically motivated adaptive systems which have been used, with good results, in function optimization. However, traditional GAs rapidly push an artificial population toward convergence. That is, all individuals in the population soon become nearly identical. Niching Methods allow genetic algorithms to maintain a population of diverse individuals. GAs that incorporate these methods are capable of locating multiple, optimal solutions within a single population. The purpose of this study is to test existing niching techniques and two methods introduced herein, bearing in mind their eventual application in nuclear reactor related problems, specially the nuclear reactor core reload one, which has multiple solutions. Tests are performed using widely known test functions and their results show that the new methods are quite promising, specially in real world problems like the nuclear reactor core reload. (author)

  11. Nuclear reactors project optimization based on neural network and genetic algorithm

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.; Schirru, Roberto; Martinez, Aquilino S.

    1997-01-01

    This work presents a prototype of a system for nuclear reactor core design optimization based on genetic algorithms and artificial neural networks. A neural network is modeled and trained in order to predict the flux and the neutron multiplication factor values based in the enrichment, network pitch and cladding thickness, with average error less than 2%. The values predicted by the neural network are used by a genetic algorithm in this heuristic search, guided by an objective function that rewards the high flux values and penalizes multiplication factors far from the required value. Associating the quick prediction - that may substitute the reactor physics calculation code - with the global optimization capacity of the genetic algorithm, it was obtained a quick and effective system for nuclear reactor core design optimization. (author). 11 refs., 8 figs., 3 tabs

  12. Optimization of in-core fuel management and control rod strategy in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1975-01-01

    An in-core fuel management problem is formulated for the equilibrium fuel cycle in an N-region nuclear reactor model. The formulation shows that the infinite multiplication factor k infinity requisite for newly charged fuel can be separated into two terms - one corresponding to the average k infinity at the end of the cycle and the other representing the direct contribution of the shuffling scheme and control rod programming. This formulation is applied to a three-region cylindrical reactor to obtain simultaneous optimization of shuffling and control rod programming. It is demonstrated that this formulation aids greatly in gaining a better understanding of the effects of changes in the shuffling scheme and control rod programming on equilibrium fuel cycle performance. (auth.)

  13. Minaret, a deterministic neutron transport solver for nuclear core calculations

    International Nuclear Information System (INIS)

    Moller, J-Y.; Lautard, J-J.

    2011-01-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  14. Minaret, a deterministic neutron transport solver for nuclear core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

    2011-07-01

    We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

  15. Reducing the risk to Mars: The gas core nuclear rocket

    International Nuclear Information System (INIS)

    Howe, S.D.; DeVolder, B.; Thode, L.; Zerkle, D.

    1998-01-01

    The next giant leap for mankind will be the human exploration of Mars. Almost certainly within the next thirty years, a human crew will brave the isolation, the radiation, and the lack of gravity to walk on and explore the Red planet. However, because the mission distances and duration will be hundreds of times greater than the lunar missions, a human crew will face much greater obstacles and a higher risk than those experienced during the Apollo program. A single solution to many of these obstacles is to dramatically decrease the mission duration by developing a high performance propulsion system. The gas-core nuclear rocket (GCNR) has the potential to be such a system. The authors have completed a comparative study of the potential impact that a GCNR could have on a manned Mars mission. The total IMLEO, transit times, and accumulated radiation dose to the crew will be compared with the NASA Design Reference Missions

  16. Optimizations of Unstructured Aerodynamics Computations for Many-core Architectures

    KAUST Repository

    Al Farhan, Mohammed Ahmed; Keyes, David E.

    2018-01-01

    involving thread and data-level parallelism. Our approach is based upon a multi-level hierarchical distribution of work and data across both the threads and the SIMD units within every hardware core. On a 64-core KNL chip, we achieve nearly 2.9x speedup

  17. Nuclear many-body problem with repulsive hard core interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haddad, L M

    1965-07-01

    The nuclear many-body problem is considered using the perturbation-theoretic approach of Brueckner and collaborators. This approach is outlined with particular attention paid to the graphical representation of the terms in the perturbation expansion. The problem is transformed to centre-of-mass coordinates in configuration space and difficulties involved in ordinary methods of solution of the resulting equation are discussed. A new technique, the 'reference spectrum method', devised by Bethe, Brandow and Petschek in an attempt to simplify the numerical work in presented. The basic equations are derived in this approximation and considering the repulsive hard core part of the interaction only, the effective mass is calculated at high momentum (using the same energy spectrum for both 'particle' and 'hole' states). The result of 0.87m is in agreement with that of Bethe et al. A more complete treatment using the reference spectrum method in introduced and a self-consistent set of equations is established for the reference spectrum parameters again for the case of hard core repulsions. (author)

  18. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  19. Development of an automated core model for nuclear reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1998-01-01

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input

  20. Optimization design for drain to nuclear power condenser

    International Nuclear Information System (INIS)

    Ding Jiapeng; Jiang Chengren

    2010-01-01

    Characters and varieties of drain to nuclear power condenser are discussed in this paper. Take the main steam system of a nuclear power as an example, normal and detailed optimization design are introduced, related expatiate are used as a reference for the drain of other systems. According to the characters of nuclear power instant operation, the influence and needed actions related with the optimization design are also analyzed. Based on the above research, the scheme has been carried out in a nuclear power station and safety for the condenser operation of the nuclear power has been improved largely. (authors)

  1. A genetic algorithm solution for combinatorial problems - the nuclear core reload example

    Energy Technology Data Exchange (ETDEWEB)

    Schirru, R.; Silva, F.C. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia; Pereira, C.M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Chapot, J.L.C. [FURNAS, Rio de Janeiro, RJ (Brazil)

    1997-12-01

    This paper presents a solution to Traveling Salesman Problem based upon genetic algorithms (GA), using the classic crossover, but avoiding the feasibility problem in offspring individuals, allowing the natural evolution of the GA without introduction of heuristics in the genetic crossover operator. The genetic model presented, that we call the List Model (LM) is based on the encoding and decoding genotype in the way to always generate a phenotype that has a valid structure, over which will be applied the fitness, represented by the total distance. The main purpose of this work was to develop the basis for a new genetic model to be used in the reload of nuclear core of a PWR. In a generic way, this problem can be interpreted as a a search of the optimal combination of N different fuel elements in N nuclear core `holes`, where each combination or load pattern, determines the neutron flux shape and its associate peak factor. The goal is to find out the load pattern that minimizes the peak factor and consequently maximize the useful life of the nuclear fuel. The GA with the List Model was applied to the Angra-1 PWR reload problem and the results are remarkably better than the ones used in the last fuel cycle. (author). 12 refs., 3 figs., 2 tabs.

  2. Optimal estimation and control in nuclear power plants

    International Nuclear Information System (INIS)

    Purviance, J.E.; Tylee, J.L.

    1982-08-01

    Optimal estimation and control theories offer the potential for more precise control and diagnosis of nuclear power plants. The important element of these theories is that a mathematical plant model is used in conjunction with the actual plant data to optimize some performance criteria. These criteria involve important plant variables and incorporate a sense of the desired plant performance. Several applications of optimal estimation and control to nuclear systems are discussed

  3. Nuclear equation of state for core-collapse supernova simulations with realistic nuclear forces

    Energy Technology Data Exchange (ETDEWEB)

    Togashi, H., E-mail: hajime.togashi@riken.jp [Nishina Center for Accelerator-Based Science, Institute of Physical and Chemical Research (RIKEN), 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Nakazato, K. [Faculty of Arts and Science, Kyushu University, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Takehara, Y.; Yamamuro, S.; Suzuki, H. [Department of Physics, Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan); Takano, M. [Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan); Department of Pure and Applied Physics, Graduate School of Advanced Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku-ku, Tokyo 169-8555 (Japan)

    2017-05-15

    A new table of the nuclear equation of state (EOS) based on realistic nuclear potentials is constructed for core-collapse supernova numerical simulations. Adopting the EOS of uniform nuclear matter constructed by two of the present authors with the cluster variational method starting from the Argonne v18 and Urbana IX nuclear potentials, the Thomas–Fermi calculation is performed to obtain the minimized free energy of a Wigner–Seitz cell in non-uniform nuclear matter. As a preparation for the Thomas–Fermi calculation, the EOS of uniform nuclear matter is modified so as to remove the effects of deuteron cluster formation in uniform matter at low densities. Mixing of alpha particles is also taken into account following the procedure used by Shen et al. (1998, 2011). The critical densities with respect to the phase transition from non-uniform to uniform phase with the present EOS are slightly higher than those with the Shen EOS at small proton fractions. The critical temperature with respect to the liquid–gas phase transition decreases with the proton fraction in a more gradual manner than in the Shen EOS. Furthermore, the mass and proton numbers of nuclides appearing in non-uniform nuclear matter with small proton fractions are larger than those of the Shen EOS. These results are consequences of the fact that the density derivative coefficient of the symmetry energy of our EOS is smaller than that of the Shen EOS.

  4. Multiobjective optimization for design of multifunctional sandwich panel heat pipes with micro-architected truss cores

    International Nuclear Information System (INIS)

    Roper, Christopher S.

    2011-01-01

    A micro-architected multifunctional structure, a sandwich panel heat pipe with a micro-scale truss core and arterial wick, is modeled and optimized. To characterize multiple functionalities, objective equations are formulated for density, compressive modulus, compressive strength, and maximum heat flux. Multiobjective optimization is used to determine the Pareto-optimal design surfaces, which consist of hundreds of individually optimized designs. The Pareto-optimal surfaces for different working fluids (water, ethanol, and perfluoro(methylcyclohexane)) as well as different micro-scale truss core materials (metal, ceramic, and polymer) are determined and compared. Examination of the Pareto fronts allows comparison of the trade-offs between density, compressive stiffness, compressive strength, and maximum heat flux in the design of multifunctional sandwich panel heat pipes with micro-scale truss cores. Heat fluxes up to 3.0 MW/m 2 are predicted for silicon carbide truss core heat pipes with water as the working fluid.

  5. State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

    2017-02-15

    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.

  6. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  7. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto

    2017-01-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10"1"3 combinations and 10"1"1 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  8. Proposal optimization in nuclear accident emergency decision based on IAHP

    International Nuclear Information System (INIS)

    Xin Jing

    2007-01-01

    On the basis of establishing the multi-layer structure of nuclear accident emergency decision, several decision objectives are synthetically analyzed, and an optimization model of decision proposals for nuclear accident emergency based on interval analytic hierarchy process is proposed in the paper. The model makes comparisons among several emergency decision proposals quantified, and the optimum proposal is selected out, which solved the uncertain and fuzzy decision problem of judgments by experts' experiences in nuclear accidents emergency decision. Case study shows that the optimization result is much more reasonable, objective and reliable than subjective judgments, and it could be decision references for nuclear accident emergency. (authors)

  9. Core optimization studies for a small heating reactor

    International Nuclear Information System (INIS)

    Galperin, A.

    1986-11-01

    Small heating reactor cores are characterized by a high contribution of the leakage to the neutron balance and by a large power density variation in the axial direction. A limited number of positions is available for the control rods, which are necessary to satisfy overall reactivity requirements subject to a safety related constraint on the maximum worth of each rod. Design approaches aimed to improve safety and fuel utilization performance of the core include separation of the cooling and moderating functions of the water with the core in order to reduce hot-to-cold reactivity shift and judicious application of the axial Gd zoning aimed to improve the discharge burnup distribution. Several design options are analyzed indicating a satisfactory solution of the axial burnup distribution problem. The feasibility of the control rod system including zircaloy, stainless steel, natural boron and possibly enriched boron rods is demonstrated. A preliminary analysis indicates directions for further improvements of the core performance by an additional reduction of the hot-to-cold reactivity shift and by a reduction of the depletion reactivity swing adopting a higher gadolinium concentration in the fuel or a two-batch fuel management scheme. (author)

  10. Finding optimal measures of core inflation in the Kyrgyz Republic

    Czech Academy of Sciences Publication Activity Database

    Uzagalieva, Ainura

    -, č. 261 (2005), s. 1-37 ISSN 1211-3298 Institutional research plan: CEZ:AV0Z70850503 Keywords : Kyrgyz Republic * core inflation * monetary policy Subject RIV: AH - Economics http://www.cerge-ei.cz/pdf/wp/Wp261.pdf

  11. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  12. Application of colony complex algorithm to nuclear component optimization design

    International Nuclear Information System (INIS)

    Yan Changqi; Li Guijing; Wang Jianjun

    2014-01-01

    Complex algorithm (CA) has got popular application to the region of nuclear engineering. In connection with the specific features of the application of traditional complex algorithm (TCA) to the optimization design in engineering structures, an improved method, colony complex algorithm (CCA), was developed based on the optimal combination of many complexes, in which the disadvantages of TCA were overcame. The optimized results of benchmark function show that CCA has better optimizing performance than TCA. CCA was applied to the high-pressure heater optimization design, and the optimization effect is obvious. (authors)

  13. Optimal Design and Analysis of the Stepped Core for Wireless Power Transfer Systems

    Directory of Open Access Journals (Sweden)

    Xiu Zhang

    2016-01-01

    Full Text Available The key of wireless power transfer technology rests on finding the most suitable means to improve the efficiency of the system. The wireless power transfer system applied in implantable medical devices can reduce the patients’ physical and economic burden because it will achieve charging in vitro. For a deep brain stimulator, in this paper, the transmitter coil is designed and optimized. According to the previous research results, the coils with ferrite core can improve the performance of the wireless power transfer system. Compared with the normal ferrite core, the stepped core can produce more uniform magnetic flux density. In this paper, the finite element method (FEM is used to analyze the system. The simulation results indicate that the core loss generated in the optimal stepped ferrite core can reduce about 10% compared with the normal ferrite core, and the efficiency of the wireless power transfer system can be increased significantly.

  14. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Directory of Open Access Journals (Sweden)

    Andrea Cerutti

    Full Text Available Hepatitis C virus (HCV infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS, but no nuclear export signal (NES has yet been identified.We show here that the aa(109-133 region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126 in the identified NES or in the sequence encoding the mature core aa(1-173 significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  15. Identification of a functional, CRM-1-dependent nuclear export signal in hepatitis C virus core protein.

    Science.gov (United States)

    Cerutti, Andrea; Maillard, Patrick; Minisini, Rosalba; Vidalain, Pierre-Olivier; Roohvand, Farzin; Pecheur, Eve-Isabelle; Pirisi, Mario; Budkowska, Agata

    2011-01-01

    Hepatitis C virus (HCV) infection is a major cause of chronic liver disease worldwide. HCV core protein is involved in nucleocapsid formation, but it also interacts with multiple cytoplasmic and nuclear molecules and plays a crucial role in the development of liver disease and hepatocarcinogenesis. The core protein is found mostly in the cytoplasm during HCV infection, but also in the nucleus in patients with hepatocarcinoma and in core-transgenic mice. HCV core contains nuclear localization signals (NLS), but no nuclear export signal (NES) has yet been identified.We show here that the aa(109-133) region directs the translocation of core from the nucleus to the cytoplasm by the CRM-1-mediated nuclear export pathway. Mutagenesis of the three hydrophobic residues (L119, I123 and L126) in the identified NES or in the sequence encoding the mature core aa(1-173) significantly enhanced the nuclear localisation of the corresponding proteins in transfected Huh7 cells. Both the NES and the adjacent hydrophobic sequence in domain II of core were required to maintain the core protein or its fragments in the cytoplasmic compartment. Electron microscopy studies of the JFH1 replication model demonstrated that core was translocated into the nucleus a few minutes after the virus entered the cell. The blockade of nucleocytoplasmic export by leptomycin B treatment early in infection led to the detection of core protein in the nucleus by confocal microscopy and coincided with a decrease in virus replication.Our data suggest that the functional NLS and NES direct HCV core protein shuttling between the cytoplasmic and nuclear compartments, with at least some core protein transported to the nucleus. These new properties of HCV core may be essential for virus multiplication and interaction with nuclear molecules, influence cell signaling and the pathogenesis of HCV infection.

  16. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  17. Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

  18. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    International Nuclear Information System (INIS)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable

  19. Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

    International Nuclear Information System (INIS)

    Prelas, M.A.; Boody, F.P.; Zediker, M.S.

    1987-01-01

    An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

  20. Seismic response of a block-type nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Bennett, J.G.; Merson, J.L.

    1976-05-01

    An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

  1. Mathematical optimization of incore nuclear fuel management decisions: Status and trends

    International Nuclear Information System (INIS)

    Turinsky, P.J.

    1999-01-01

    Nuclear fuel management involves making decisions about the number of fresh assemblies to purchase and their Attributes (e.g. enrichment and burnable poison loading), burnt fuel to reinsert, location of the assemblies in the core (i.e. loading pattern (LP)), and insertion of control rods as a function of cycle exposure (i.e. control rod pattern (CRP)). The out-of-core and incore nuclear fuel management problems denote an artificial separation of decisions to simplify the decisionmaking. The out-of-core problem involves multicycle analysis so that levelized fuel cycle cost can be evaluated; whereas, the incore problem normally involves single cycle analysis. Decision variables for the incore problem normally include all of the above noted decisions with the exception of the number of fresh assemblies, which is restricted by discharge burnup limits and therefore involves multicycle considerations. This paper reports on the progress that is being made in addressing the incore nuclear fuel management problem utilizing formal mathematical optimization methods. Advances in utilizing the Simulating Annealing, Genetic Algorithm and Tabu Search methods, with applications to pressurized and boiling water reactor incore optimization problem, will be reviewed. Recent work on the addition of multiobjective optimization capability to aide the decision maker, and utilization of heuristic rules and incorporation of parallel algorithms to increase computational efficiency, will be discussed. (orig.) [de

  2. Optimization method development of the core characteristics of a fast reactor in order to explore possible high performance solutions (a solution being a consistent set of fuel, core, system and safety)

    International Nuclear Information System (INIS)

    Ingremeau, J.-J.X.

    2011-01-01

    In the study of any new nuclear reactor, the design of the core is an important step. However designing and optimising a reactor core is quite complex as it involves neutronics, thermal-hydraulics and fuel thermomechanics and usually design of such a system is achieved through an iterative process, involving several different disciplines. In order to solve quickly such a multi-disciplinary system, while observing the appropriate constraints, a new approach has been developed to optimise both the core performance (in-cycle Pu inventory, fuel burn-up, etc...) and the core safety characteristics (safety estimators) of a Fast Neutron Reactor. This new approach, called FARM (Fast Reactor Methodology) uses analytical models and interpolations (Meta-models) from CEA reference codes for neutronics, thermal-hydraulics and fuel behaviour, which are coupled to automatically design a core based on several optimization variables. This global core model is then linked to a genetic algorithm and used to explore and optimise new core designs with improved performance. Consideration has also been given to which parameters can be best used to define the core performance and how safety can be taken into account.This new approach has been used to optimize the design of three concepts of Gas cooled Fast Reactor (GFR). For the first one, using a SiC/SiCf-cladded carbide-fuelled helium-bonded pin, the results demonstrate that the CEA reference core obtained with the traditional iterative method was an optimal core, but among many other possibilities (that is to say on the Pareto front). The optimization also found several other cores which exhibit some improved features at the expense of other safety or performance estimators. An evolution of this concept using a 'buffer', a new technology being developed at CEA, has hence been introduced in FARM. The FARM optimisation produced several core designs using this technology, and estimated their performance. The results obtained show that

  3. Using package MESSAGE for optimization studies of nuclear power structures

    International Nuclear Information System (INIS)

    Andrianov, A.A.; Fedorova, E.V.; Korobejnikov, V.V.; Poplavskaya, E.V.; Rachkova, E.N.

    2010-01-01

    The results of optimization research of Russia nuclear power strategies, obtained for different assumptions concerning availability of natural uranium resources were presented. The ability of energy planning package MESSAGE (Model for Energy Supply Strategy Alternatives and their General Environmental Impacts) application for elaborating breeding parameters requirements of fast sodium reactors and assessing the required scale of nuclear fuel cycle enterprises development was demonstrated [ru

  4. Optimal Physical Protection against Nuclear Terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Doyoung; Kim, ChangLak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    There is no attempt with nuclear weapons to attack any places for terror or military victory since the atomic bombs dropped in Hiroshima and Nagasaki. People have obviously experienced horrible destructive power of nuclear weapons and continuously remembered a terrible tragedy, lots of organizations and experts express their concerns about the nuclear terrorism and try to interchange opinions for prevention of deadly weapons. The purpose of this paper is to provide the information of nuclear terrorism and what the potential risk of Republic of Korea is and how to do the efficient physical protection. Terror is from the old French terreur, which is derived from Latin verb terror meaning 'great fear'. This is a policy to suppress political opponents through using violence and repression. Many scholars have been proposed, there is no consensus definition of the term 'terrorism.' In 1988, a proposed academic consensus definition: 'Terrorism is an anxiety-inspiring method of repeated violent action, employed by (semi-) clandestine individual, group or state actors, for idiosyncratic, criminal or political reasons, whereby - in contrast to assassination - the direct targets of violence are not the main targets. The immediate human victims of violence are generally chosen randomly (targets of opportunity) or selectively (representative or symbolic targets) from a target population, and serve as message generators. These attacks showed that particular terrorists groups sought to cause heavy casualties and extreme terrorists were spontaneously prepared to make sacrifices for completion of that ultimate goal. Creation of nuclear weapons was like opening Pandora's box. Barack Obama has called nuclear terrorism 'the greatest danger we face'. Nuclear terror is one of the lethal risks. Using nuclear weapons or materials from terrorist groups is a fatal catastrophe to a targeting state though there is no accident similar like that. South

  5. Optimal Physical Protection against Nuclear Terrorism

    International Nuclear Information System (INIS)

    Lee, Doyoung; Kim, ChangLak

    2014-01-01

    There is no attempt with nuclear weapons to attack any places for terror or military victory since the atomic bombs dropped in Hiroshima and Nagasaki. People have obviously experienced horrible destructive power of nuclear weapons and continuously remembered a terrible tragedy, lots of organizations and experts express their concerns about the nuclear terrorism and try to interchange opinions for prevention of deadly weapons. The purpose of this paper is to provide the information of nuclear terrorism and what the potential risk of Republic of Korea is and how to do the efficient physical protection. Terror is from the old French terreur, which is derived from Latin verb terror meaning 'great fear'. This is a policy to suppress political opponents through using violence and repression. Many scholars have been proposed, there is no consensus definition of the term 'terrorism.' In 1988, a proposed academic consensus definition: 'Terrorism is an anxiety-inspiring method of repeated violent action, employed by (semi-) clandestine individual, group or state actors, for idiosyncratic, criminal or political reasons, whereby - in contrast to assassination - the direct targets of violence are not the main targets. The immediate human victims of violence are generally chosen randomly (targets of opportunity) or selectively (representative or symbolic targets) from a target population, and serve as message generators. These attacks showed that particular terrorists groups sought to cause heavy casualties and extreme terrorists were spontaneously prepared to make sacrifices for completion of that ultimate goal. Creation of nuclear weapons was like opening Pandora's box. Barack Obama has called nuclear terrorism 'the greatest danger we face'. Nuclear terror is one of the lethal risks. Using nuclear weapons or materials from terrorist groups is a fatal catastrophe to a targeting state though there is no accident similar like that. South

  6. Maintenance optimization in nuclear power plants through genetic algorithms

    International Nuclear Information System (INIS)

    Munoz, A.; Martorell, S.; Serradell, V.

    1999-01-01

    Establishing suitable scheduled maintenance tasks leads to optimizing the reliability of nuclear power plant safety systems. The articles addresses this subject, whilst endeavoring to tackle an overall optimization process for component availability and safety systems through the use of genetic algorithms. (Author) 20 refs

  7. Methods optimization for the first time core critical

    International Nuclear Information System (INIS)

    Yan Liang

    2014-01-01

    The PWR reactor core commissioning programs the content of the first critical reactor physics experiment, and describes thc physical test method. However, all the methods arc not exactly the same but efficient. This article aims to enhance the reactor for the first time in the process of critical safety, shorten the overall time of critical physical test for the first time, and improve the integrity of critical physical test data for the first time and accuracy, eventually to improve the operation of the plant economic benefit adopting sectional dilution, power feedback for Doppler point improvement of physical test methods, and so on. (author)

  8. Optimization of the commissioning period of nuclear power plant

    International Nuclear Information System (INIS)

    Hou Ganglian; Li Chunyue

    2014-01-01

    Due to current equipment manufacture capacity, construction experience and other factors, commissioning of nuclear power projects was used to be postponed, which could lead to delay of the whole project. Based on the actual situation, optimization of commissioning period and its logic could be an effective way to improve this situation to some extent. Based on previous practice and experience in the schedule management for the commissioning nuclear power projects, this paper analyzes and discusses the characteristics of make commissioning plan and the difficulties of program implementation and strategies of commissioning plan optimization, discusses and presents ways of dynamic plan adjustment and optimization at the vision of entire project, synthesizes the methods of time management through commissioning itself, interface and management, expounds measures for the timing and optimization of commissioning schedule and commissioning period, and sums up the ways of optimization of commissioning period, improving management capabilities and control of optimization principles. (authors)

  9. Optimization of core reload design for low-leakage fuel management in pressurized water reactors

    International Nuclear Information System (INIS)

    Kim, Y.J.; Downar, T.J.; Sesonske, A.

    1987-01-01

    A method was developed to optimize pressurized water reactor low-leakage core reload designs that features the decoupling and sequential optimization of the fuel arrangement and control problems. The two-stage optimization process provides the maximum cycle length for a given fresh fuel loading subject to power peaking constraints. In the first stage, a best fuel arrangement is determined at the end of cycle (EOC) in the absence of all control poisons by employing a direct search method. The constant power, Haling depletion is used to provide the cycle length and EOC power peaking for each candidate core fuel arrangement. In the second stage, the core control poison requirements to meet the core peaking constraints throughout the cycle are determined using an approximate nonlinear programming technique

  10. Robust optimal self tuning regulator of nuclear reactors

    International Nuclear Information System (INIS)

    Nouri Khajavi, M.; Menhaj, M.B.; Ghofrani, M.B.

    2000-01-01

    Nuclear power reactors are, in nature nonlinear and time varying. These characteristics must be considered, if large power variations occur in their working regime. In this paper a robust optimal self-tuning regulator for regulating the power of a nuclear reactor has been designed and simulated. The proposed controller is capable of regulating power levels in a wide power range (10% to 100% power levels). The controller achieves a fast and good transient response. The simulation results show that the proposed controller outperforms the fixed optimal control recently cited in the literature for nuclear power plants

  11. Computer realization of an algorithm for determining the optimal arrangement of a fast power reactor core with hexagonal assemblies

    International Nuclear Information System (INIS)

    Karpov, V.A.; Rybnikov, A.F.

    1983-01-01

    An algorithm for solving the problems associated with fast nuclear reactor computer-aided design is suggested. Formulation of the discrete optimization problem dealing with chosing of the first loading arrangement, determination of the control element functional purpose and the order of their rearrangement during reactor operation as well as the choice of operations for core reloading is given. An algorithm for computerized solutions of the mentioned optimization problem based on variational methods relized in the form of the DESIGN program complex written in FORTRAN for the BEhSM-6 computer is proposed. A fast-response program for solving the diffusion equations of two-dimensional reactor permitting to obtain the optimization problem solution at reasonable period of time is developed to conduct necessary neutron-physical calculations for the reactor in hexagonal geometry. The DESIGN program can be included into a computer-aided design system for automation of the procedure of determining the fast power reactor core arrangement. Application of the DESIGN program permits to avoid the routine calculations on substantiation of neutron-physical and thermal-hydraulic characteristics of the reactor core that releases operators from essential waste of time and increases efficiency of their work

  12. A discrete optimization method for nuclear fuel management

    International Nuclear Information System (INIS)

    Argaud, J.P.

    1993-01-01

    Nuclear fuel management can be seen as a large discrete optimization problem under constraints, and optimization methods on such problems are numerically costly. After an introduction of the main aspects of nuclear fuel management, this paper presents a new way to treat the combinatorial problem by using information included in the gradient of optimized cost function. A new search process idea is to choose, by direct observation of the gradient, the more interesting changes in fuel loading patterns. An example is then developed to illustrate an operating mode of the method. Finally, connections with classical simulated annealing and genetic algorithms are described as an attempt to improve search processes. 16 refs., 2 figs

  13. On the optimal design of shell and tube heat exchanger for nuclear applications

    International Nuclear Information System (INIS)

    Abd Rabbo, F.M.M.; Fatb Allab, A.; El-Fawal, M.

    1997-01-01

    In nuclear industry, heat exchanger plays an important role in the transfer of heat from reactor core, where heat is generated, to the ultimate heat sink UHS, and then is dissipated. The actual design of heat exchanger not only relies on thermohydraulic considerations but also on economical aspects and radiological safety considerations. For optimal design of heat exchanger for a specific application a compromise should be made for determining the important factors affecting the design. In this paper, an optimization model is presented for shell and tube heat exchanger, which could be considered as a tool for computer aided design. A case study is presented to explore the present adopted model. 3 figs

  14. Characteristic statistic algorithm (CSA) for in-core loading pattern optimization

    International Nuclear Information System (INIS)

    Liu Zhihong; Hu Yongming; Shi Gong

    2007-01-01

    To solve the problem of PWR in-core loading pattern optimization, a more suitable global optimization algorithm, i.e., Characteristic statistic algorithm (CSA), is used. The searching process of this algorithm and how to apply it to this problem are presented. Loading pattern optimization code SCYCLE is developed. Two different problems on real PWR models are calculated and the results are compared with other algorithms. It is shown that SCYCLE has high efficiency and good global performance on this problem. (authors)

  15. Particle swarm optimization with random keys applied to the nuclear reactor reload problem

    Energy Technology Data Exchange (ETDEWEB)

    Meneses, Anderson Alvarenga de Moura [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear; Fundacao Educacional de Macae (FUNEMAC), RJ (Brazil). Faculdade Professor Miguel Angelo da Silva Santos; Machado, Marcelo Dornellas; Medeiros, Jose Antonio Carlos Canedo; Schirru, Roberto [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: ameneses@con.ufrj.br; marcelo@lmp.ufrj.br; canedo@lmp.ufrj.br; schirru@lmp.ufrj.br

    2007-07-01

    In 1995, Kennedy and Eberhart presented the Particle Swarm Optimization (PSO), an Artificial Intelligence metaheuristic technique to optimize non-linear continuous functions. The concept of Swarm Intelligence is based on the socials aspects of intelligence, it means, the ability of individuals to learn with their own experience in a group as well as to take advantage of the performance of other individuals. Some PSO models for discrete search spaces have been developed for combinatorial optimization, although none of them presented satisfactory results to optimize a combinatorial problem as the nuclear reactor fuel reloading problem (NRFRP). In this sense, we developed the Particle Swarm Optimization with Random Keys (PSORK) in previous research to solve Combinatorial Problems. Experiences demonstrated that PSORK performed comparable to or better than other techniques. Thus, PSORK metaheuristic is being applied in optimization studies of the NRFRP for Angra 1 Nuclear Power Plant. Results will be compared with Genetic Algorithms and the manual method provided by a specialist. In this experience, the problem is being modeled for an eight-core symmetry and three-dimensional geometry, aiming at the minimization of the Nuclear Enthalpy Power Peaking Factor as well as the maximization of the cycle length. (author)

  16. Particle swarm optimization with random keys applied to the nuclear reactor reload problem

    International Nuclear Information System (INIS)

    Meneses, Anderson Alvarenga de Moura; Fundacao Educacional de Macae; Machado, Marcelo Dornellas; Medeiros, Jose Antonio Carlos Canedo; Schirru, Roberto

    2007-01-01

    In 1995, Kennedy and Eberhart presented the Particle Swarm Optimization (PSO), an Artificial Intelligence metaheuristic technique to optimize non-linear continuous functions. The concept of Swarm Intelligence is based on the socials aspects of intelligence, it means, the ability of individuals to learn with their own experience in a group as well as to take advantage of the performance of other individuals. Some PSO models for discrete search spaces have been developed for combinatorial optimization, although none of them presented satisfactory results to optimize a combinatorial problem as the nuclear reactor fuel reloading problem (NRFRP). In this sense, we developed the Particle Swarm Optimization with Random Keys (PSORK) in previous research to solve Combinatorial Problems. Experiences demonstrated that PSORK performed comparable to or better than other techniques. Thus, PSORK metaheuristic is being applied in optimization studies of the NRFRP for Angra 1 Nuclear Power Plant. Results will be compared with Genetic Algorithms and the manual method provided by a specialist. In this experience, the problem is being modeled for an eight-core symmetry and three-dimensional geometry, aiming at the minimization of the Nuclear Enthalpy Power Peaking Factor as well as the maximization of the cycle length. (author)

  17. A hybrid method for in-core optimization of pressurized water reactor reload core design

    International Nuclear Information System (INIS)

    Stevens, J.G.

    1995-05-01

    The objective of this research is the development of an accurate, practical, and robust method for optimization of the design of loading patterns for pressurized water reactors, a nonlinear, non-convex, integer optimization problem. The many logical constraints which may be applied during the design process are modeled herein by a network construction upon which performance objectives and safety constraints from reactor physics calculations are optimized. This thesis presents the synthesis of the strengths of previous algorithms developed for reload design optimization and extension of robustness through development of a hybrid liberated search algorithm. Development of three independent methods for reload design optimization is presented: random direct search for local improvement, liberated search by simulated annealing, and deterministic search for local improvement via successive linear assignment by branch and bound. Comparative application of the methods to a variety of problems is discussed, including an exhaustive enumeration benchmark created to allow comparison of search results to a known global optimum for a large scale problem. While direct search and determinism are shown to be capable of finding improvement, only the liberation of simulated annealing is found to perform robustly in the non-convex design spaces. The hybrid method SHAMAN is presented. The algorithm applies: determinism to shuffle an initial solution for satisfaction of heuristics and symmetry; liberated search through simulated annealing with a bounds cooling constraint treatment; and search bias through relational heuristics for the application of engineering judgment. The accuracy, practicality, and robustness of the SHAMAN algorithm is demonstrated through application to a variety of reload loading pattern optimization problems

  18. Nuclear fuel management optimization using adaptive evolutionary algorithms with heuristics

    International Nuclear Information System (INIS)

    Axmann, J.K.; Van de Velde, A.

    1996-01-01

    Adaptive Evolutionary Algorithms in combination with expert knowledge encoded in heuristics have proved to be a robust and powerful optimization method for the design of optimized PWR fuel loading pattern. Simple parallel algorithmic structures coupled with a low amount of communications between computer processor units in use makes it possible for workstation clusters to be employed efficiently. The extension of classic evolution strategies not only by new and alternative methods but also by the inclusion of heuristics with effects on the exchange probabilities of the fuel assemblies at specific core positions leads to the RELOPAT optimization code of the Technical University of Braunschweig. In combination with the new, neutron-physical 3D nodal core simulator PRISM developed by SIEMENS the PRIMO loading pattern optimization system has been designed. Highly promising results in the recalculation of known reload plans for German PWR's new lead to a commercially usable program. (author)

  19. Optimization in the scale of nuclear power generation and the economy of nuclear power

    International Nuclear Information System (INIS)

    Suzuki, Toshiharu

    1983-01-01

    In the not too distant future, the economy of nuclear power will have to be restudied. Various conditions and circumstances supporting this economy of nuclear power tend to change, such as the decrease in power demand and supply, the diversification in base load supply sources, etc. The fragility in the economic advantage of nuclear power may thus be revealed. In the above connection, on the basis of the future outlook of the scale of nuclear power generation, that is, the further reduction of the current nuclear power program, and of the corresponding supply and demand of nuclear fuel cycle quantities, the aspect of the economic advantage of nuclear power was examined, for the purpose of optimizing the future scale of nuclear power generation (the downward revision of the scale, the establishment of the schedule of nuclear fuel cycle the stagnation of power demand and nuclear power generation costs). (Mori, K.)

  20. Molten core material holding device in a nuclear reactor

    International Nuclear Information System (INIS)

    Nakamura, Hisashi; Tanaka, Nobuo; Takahashi, Katsuro.

    1985-01-01

    Purpose: To improve the function of cooling to hold molten core materials in a molten core material holding device. Constitution: Plenum structures are formed into a pan-like configuration, in which liners made of metal having high melting point and relatively high heat conductivity such as tantalum, tungsten, rhenium or alloys thereof are integrally appended to hold and directly cool the molten reactor core materials. Further, a plurality of heat pipes, passing through the plenum structures, facing the cooling portion thereof to the coolants at the outer side and immersing the heating portion into the molten core materials fallen to deposit in the inner liners are disposed radially. Furthermore, heat pipes embodded in the plenum structure are disposed in the same manner below the liners. Thus, the plenum structures and the molten reactor core materials can be cooled at a high efficiency. (Seki, T.)

  1. Nuclear-fuel-cycle education: Module 5. In-core fuel management

    International Nuclear Information System (INIS)

    Levine, S.H.

    1980-07-01

    The purpose of this project was to develop a series of educational modules for use in nuclear-fuel-cycle education. These modules are designed for use in a traditional classroom setting by lectures or in a self-paced, personalized system of instruction. This module on in-core fuel management contains information on computational methods and theory; in-core fuel management using the Virginia Polytechnic Institute and State University computer modules; pressurized water reactor in-core fuel management; boiling water reactor in-core fuel management; and in-core fuel management for gas-cooled and fast reactors

  2. Construction of Core Collections Suitable for Association Mapping to Optimize Use of Mediterranean Olive (Olea europaea L.) Genetic Resources

    Science.gov (United States)

    El Bakkali, Ahmed; Haouane, Hicham; Moukhli, Abdelmajid; Costes, Evelyne; Van Damme, Patrick; Khadari, Bouchaib

    2013-01-01

    Phenotypic characterisation of germplasm collections is a decisive step towards association mapping analyses, but it is particularly expensive and tedious for woody perennial plant species. Characterisation could be more efficient if focused on a reasonably sized subset of accessions, or so-called core collection (CC), reflecting the geographic origin and variability of the germplasm. The questions that arise concern the sample size to use and genetic parameters that should be optimized in a core collection to make it suitable for association mapping. Here we investigated these questions in olive (Olea europaea L.), a perennial fruit species. By testing different sampling methods and sizes in a worldwide olive germplasm bank (OWGB Marrakech, Morocco) containing 502 unique genotypes characterized by nuclear and plastid loci, a two-step sampling method was proposed. The Shannon-Weaver diversity index was found to be the best criterion to be maximized in the first step using the Core Hunter program. A primary core collection of 50 entries (CC50) was defined that captured more than 80% of the diversity. This latter was subsequently used as a kernel with the Mstrat program to capture the remaining diversity. 200 core collections of 94 entries (CC94) were thus built for flexibility in the choice of varieties to be studied. Most entries of both core collections (CC50 and CC94) were revealed to be unrelated due to the low kinship coefficient, whereas a genetic structure spanning the eastern and western/central Mediterranean regions was noted. Linkage disequilibrium was observed in CC94 which was mainly explained by a genetic structure effect as noted for OWGB Marrakech. Since they reflect the geographic origin and diversity of olive germplasm and are of reasonable size, both core collections will be of major interest to develop long-term association studies and thus enhance genomic selection in olive species. PMID:23667437

  3. The effects of radiation on aluminium alloys in the core of energy nuclear reactors

    International Nuclear Information System (INIS)

    Petrossian, V.G.

    1995-01-01

    One of the attractive directions in the worldwide practice of nuclear installations is the replacement of expensive zirconium alloy with more cheap materials, particularly aluminium allo. For Heat Supply Nuclear Plants (HSNP) with approximately 473 K core temperatures, the use of heat-resistant aluminium alloys seems to be reasonable. The present work is concerned with the studies on radiation effects on aluminium alloy, and interaction between the alloy and coolant in the reactor core. (author). 2 refs., 3 figs., 1 tab

  4. Accelerating the SCE-UA Global Optimization Method Based on Multi-Core CPU and Many-Core GPU

    Directory of Open Access Journals (Sweden)

    Guangyuan Kan

    2016-01-01

    Full Text Available The famous global optimization SCE-UA method, which has been widely used in the field of environmental model parameter calibration, is an effective and robust method. However, the SCE-UA method has a high computational load which prohibits the application of SCE-UA to high dimensional and complex problems. In recent years, the hardware of computer, such as multi-core CPUs and many-core GPUs, improves significantly. These much more powerful new hardware and their software ecosystems provide an opportunity to accelerate the SCE-UA method. In this paper, we proposed two parallel SCE-UA methods and implemented them on Intel multi-core CPU and NVIDIA many-core GPU by OpenMP and CUDA Fortran, respectively. The Griewank benchmark function was adopted in this paper to test and compare the performances of the serial and parallel SCE-UA methods. According to the results of the comparison, some useful advises were given to direct how to properly use the parallel SCE-UA methods.

  5. Development of advanced nuclear core analysis system applicable to various reactor types

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, aiming at development of an advanced detailed analysis system applicable to nuclear core performance analysis of various fast reactors currently considered, the concept of cross section library set was examined and the specification of library set was determined. That is to say, referring the world most advanced reactor physics analysis system ERANOS (European Reactor Analysis Optimized System) and the result of preceding research 'preparation of next generation cross section library', 900 energy groups structure, concrete cross section data to be included and the format of cross section library were defined. And we performed elaborate work revising the group cross section production system which was prepared in the preceding research. After that the revision work was completed, to confirm the capability of revised cross section production system, we produced a prototype 450 groups cross section library. And we carried out a series of bench mark tests including analysis of small fast reactors utilizing this prototype cross section library and confirmed that the prototype cross section library has sufficient accuracy for predicting core performance. Furthermore, we estimated the computer resource information such as memory size, hard disk capacity and calculation time, etc. necessary for producing 900 groups detailed cross section library. In addition, we identified problems to be solved for developing a cell calculation code installed in our detailed analysis system. (author)

  6. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  7. Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97.

    Science.gov (United States)

    Milbradt, Jens; Sonntag, Eric; Wagner, Sabrina; Strojan, Hanife; Wangen, Christina; Lenac Rovis, Tihana; Lisnic, Berislav; Jonjic, Stipan; Sticht, Heinrich; Britt, William J; Schlötzer-Schrehardt, Ursula; Marschall, Manfred

    2018-01-13

    The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC) that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV) capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

  8. Human Cytomegalovirus Nuclear Capsids Associate with the Core Nuclear Egress Complex and the Viral Protein Kinase pUL97

    Directory of Open Access Journals (Sweden)

    Jens Milbradt

    2018-01-01

    Full Text Available The nuclear phase of herpesvirus replication is regulated through the formation of regulatory multi-component protein complexes. Viral genomic replication is followed by nuclear capsid assembly, DNA encapsidation and nuclear egress. The latter has been studied intensely pointing to the formation of a viral core nuclear egress complex (NEC that recruits a multimeric assembly of viral and cellular factors for the reorganization of the nuclear envelope. To date, the mechanism of the association of human cytomegalovirus (HCMV capsids with the NEC, which in turn initiates the specific steps of nuclear capsid budding, remains undefined. Here, we provide electron microscopy-based data demonstrating the association of both nuclear capsids and NEC proteins at nuclear lamina budding sites. Specifically, immunogold labelling of the core NEC constituent pUL53 and NEC-associated viral kinase pUL97 suggested an intranuclear NEC-capsid interaction. Staining patterns with phospho-specific lamin A/C antibodies are compatible with earlier postulates of targeted capsid egress at lamina-depleted areas. Important data were provided by co-immunoprecipitation and in vitro kinase analyses using lysates from HCMV-infected cells, nuclear fractions, or infectious virions. Data strongly suggest that nuclear capsids interact with pUL53 and pUL97. Combined, the findings support a refined concept of HCMV nuclear trafficking and NEC-capsid interaction.

  9. Optimal power and distribution control for weakly-coupled-core reactor

    International Nuclear Information System (INIS)

    Oohori, Takahumi; Kaji, Ikuo

    1977-01-01

    A numerical procedure has been devised for obtaining the optimal power and distribution control for a weakly-coupled-core reactor. Several difficulties were encountered in solving this optimization problem: (1) nonlinearity of the reactor kinetics equations; (2) neutron-leakage interaction between the cores; (3) localized power changes occurring in addition to the total power changes; (4) constraints imposed on the states - e.g. reactivity, reactor period. To obviate these difficulties, use is made of the generalized Newton method to convert the problem into an iterative sequence of linear programming problems, after approximating the differential equations and the integral performance criterion by a set of discrete algebraic equations. In this procedure, a heuristic but effective method is used for deriving an initial approximation, which is then made to converge toward the optimal solution. Delayed-neutron one-group point reactor models embodying transient temperature feed-back to the reactivity are used in obtaining the kinetics equations for the weakly-coupled-core reactor. The criterion adopted for determining the optimality is a norm relevant to the deviations of neutron density from the desired trajectories or else to the time derivatives of the neutron density; uniform control intervals are prescribed. Examples are given of two coupled-core reactors with typical parameters to illustrate the results obtained with this procedure. A comparison is also made between the coupled-core reactor and the one-point reactor. (auth.)

  10. Application of nonlinear nodal diffusion generalized perturbation theory to nuclear fuel reload optimization

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.

    1995-01-01

    The determination of the family of optimum core loading patterns for pressurized water reactors (PWRs) involves the assessment of the core attributes for thousands of candidate loading patterns. For this reason, the computational capability to efficiently and accurately evaluate a reactor core's eigenvalue and power distribution versus burnup using a nodal diffusion generalized perturbation theory (GPT) model is developed. The GPT model is derived from the forward nonlinear iterative nodal expansion method (NEM) to explicitly enable the preservation of the finite difference matrix structure. This key feature considerably simplifies the mathematical formulation of NEM GPT and results in reduced memory storage and CPU time requirements versus the traditional response-matrix approach to NEM. In addition, a treatment within NEM GPT can account for localized nonlinear feedbacks, such as that due to fission product buildup and thermal-hydraulic effects. When compared with a standard nonlinear iterative NEM forward flux solve with feedbacks, the NEM GPT model can execute between 8 and 12 times faster. These developments are implemented within the PWR in-core nuclear fuel management optimization code FORMOSA-P, combining the robustness of its adaptive simulated annealing stochastic optimization algorithm with an NEM GPT neutronics model that efficiently and accurately evaluates core attributes associated with objective functions and constraints of candidate loading patterns

  11. Particle swarm as optimization tool in complex nuclear engineering problems

    International Nuclear Information System (INIS)

    Medeiros, Jose Antonio Carlos Canedo

    2005-06-01

    Due to its low computational cost, gradient-based search techniques associated to linear programming techniques are being used as optimization tools. These techniques, however, when applied to multimodal search spaces, can lead to local optima. When finding solutions for complex multimodal domains, random search techniques are being used with great efficacy. In this work we exploit the swarm optimization algorithm search power capacity as an optimization tool for the solution of complex high dimension and multimodal search spaces of nuclear problems. Due to its easy and natural representation of high dimension domains, the particle swarm optimization was applied with success for the solution of complex nuclear problems showing its efficacy in the search of solutions in high dimension and complex multimodal spaces. In one of these applications it enabled a natural and trivial solution in a way not obtained with other methods confirming the validity of its application. (author)

  12. Nuclear reactors project optimization based on neural network and genetic algorithm; Otimizacao em projetos de reatores nucleares baseada em rede neural e algoritmo genetico

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Schirru, Roberto; Martinez, Aquilino S. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia

    1997-12-01

    This work presents a prototype of a system for nuclear reactor core design optimization based on genetic algorithms and artificial neural networks. A neural network is modeled and trained in order to predict the flux and the neutron multiplication factor values based in the enrichment, network pitch and cladding thickness, with average error less than 2%. The values predicted by the neural network are used by a genetic algorithm in this heuristic search, guided by an objective function that rewards the high flux values and penalizes multiplication factors far from the required value. Associating the quick prediction - that may substitute the reactor physics calculation code - with the global optimization capacity of the genetic algorithm, it was obtained a quick and effective system for nuclear reactor core design optimization. (author). 11 refs., 8 figs., 3 tabs.

  13. Analysis of ex-core detector response measured during nuclear ship Mutsu land-loaded core critical experiment

    International Nuclear Information System (INIS)

    Itagaki, M.; Abe, J.I.; Kuribayashi, K.

    1987-01-01

    There are some cases where the ex-core neutron detector response is dependent not only on the fission source distribution in a core but also on neutron absorption in the borated water reflector. For example, an unexpectedly large response variation was measured during the nuclear ship Mutsu land-loaded core critical experiment. This large response variation is caused largely by the boron concentration change associated with the change in control rod positioning during the experiment. The conventional Crump-Lee response calculation method has been modified to take into account this boron effect. The correction factor in regard to this effect has been estimated using the one-dimensional transport code ANISN. The detector response variations obtained by means of this new calculation procedure agree well with the measured values recorded during the experiment

  14. Core power distribution measurement and data processing in Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    Zhang Hong

    1997-01-01

    For the first time in China, Daya Bay Nuclear Power Station applied the advanced technology of worldwide commercial pressurized reactors to the in-core detectors, the leading excore six-chamber instrumentation for precise axial power distribution, and the related data processing. Described in this article are the neutron flux measurement in Daya Bay Nuclear Power Station, and the detailed data processing

  15. Real-time simulation of ex-core nuclear instrumentation system

    International Nuclear Information System (INIS)

    Zhao Qiang; Zhang Zhijian; Cao Xinrong

    2005-01-01

    Real-time simulation of ex-core nuclear instrumentation system is an indispensable part of nuclear power plant (NPP) full-scope training simulator. The simulation method, which is based upon the theory of measurement, is introduced in the paper. The fitting formula between the measured data and the three-dimensional neutron flux distribution in the core is established. The fitting parameter is adjusted according to the reactor physical calculation or the experiment of power calibration. The simulation result shows that the method can simulate the ex-core neutron instrumentation system accurately in real-time and meets the needs of NPP full-scope training simulator. (authors)

  16. A study on optimization of the nuclear safety system

    International Nuclear Information System (INIS)

    Lee, Sang Hoon; Koh, Byung Joon; Kim, Jin Soo; Kim, Byoung Do; Cho, Seong Won; Kwon, Seog Kwon; Choi, Kwang Sik

    1986-12-01

    The number of nuclear facilities (nuclear power plants, research reactors, nuclear fuel facilities) under construction or in operation in Korea continues to increase and this has brought about increased importance and concerns toward nuclear safety in Korea. Also, domestic nuclear related organizations are increasingly carrying out the design/construction of nuclear power plants and the development /supply of nuclear fuels. In order to flexibly respond to these changes and to suggest direction to take, it is necessary to re-examine the current nuclear safety regulation system. This study is carried out in two stages and this report describes the results of the analysis and the assessment of the nuclear licencing system of such foreign countries as sweden and German, as the first of the two. In this regard, this study includes the analysis on the backgrounds on the choice of nuclear licensing system, the analysis on the licensing procedures, the analysis on the safety inspection system and the enforcement laws, the analysis on the structure and function of the regulatory, business and research organizations as well as the analysis on the relationship between the safety research and the regulatory duties. In this study, the German safety inspection system and the enforcement procedures and the Swedish nuclear licensing system are analyzed in detail. By comparing and assessing the finding with the current Korea Nuclear Licensing System, this study points out some reform measures of the Korean system that needs to improved. With the changing situations in mind, this study aims to develop the nuclear safety regulation system optimized for Korean situation by re-examining the current regulation system. (Author)

  17. Gas core nuclear thermal rocket engine research and development in the former USSR

    International Nuclear Information System (INIS)

    Koehlinger, M.W.; Bennett, R.G.; Motloch, C.G.; Gurfink, M.M.

    1992-09-01

    Beginning in 1957 and continuing into the mid 1970s, the USSR conducted an extensive investigation into the use of both solid and gas core nuclear thermal rocket engines for space missions. During this time the scientific and engineering. problems associated with the development of a solid core engine were resolved. At the same time research was undertaken on a gas core engine, and some of the basic engineering problems associated with the concept were investigated. At the conclusion of the program, the basic principles of the solid core concept were established. However, a prototype solid core engine was not built because no established mission required such an engine. For the gas core concept, some of the basic physical processes involved were studied both theoretically and experimentally. However, no simple method of conducting proof-of-principle tests in a neutron flux was devised. This report focuses primarily on the development of the. gas core concept in the former USSR. A variety of gas core engine system parameters and designs are presented, along with a summary discussion of the basic physical principles and limitations involved in their design. The parallel development of the solid core concept is briefly described to provide an overall perspective of the magnitude of the nuclear thermal propulsion program and a technical comparison with the gas core concept

  18. Generalized perturbation theory error control within PWR core-loading pattern optimization

    International Nuclear Information System (INIS)

    Imbriani, J.S.; Turinsky, P.J.; Kropaczek, D.J.

    1995-01-01

    The fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for PWR reactors. The code couples the optimization technique of simulated annealing (SA) with a generalized perturbation theory (GPT) model for evaluating core physics characteristics. To ensure the accuracy of the GPT predictions, as well as to maximize the efficient of the SA search, a GPT error control method has been developed

  19. Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

    2008-07-01

    A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

  20. Nuclear cardiology core syllabus of the European Association of Cardiovascular Imaging (EACVI).

    Science.gov (United States)

    Gimelli, Alessia; Neglia, Danilo; Schindler, Thomas H; Cosyns, Bernard; Lancellotti, Patrizio; Kitsiou, Anastasia

    2015-04-01

    The European Association of Cardiovascular Imaging (EACVI) Core Syllabus for Nuclear Cardiology is now available online. The syllabus lists key elements of knowledge in nuclear cardiology. It represents a framework for the development of training curricula and provides expected knowledge-based learning outcomes to the nuclear cardiology trainees. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2015. For permissions please email: journals.permissions@oup.com.

  1. Third-generation site characterization: Cryogenic core collection, nuclear magnetic resonance, and electrical resistivity

    Science.gov (United States)

    Kiaalhosseini, Saeed

    In modern contaminant hydrology, management of contaminated sites requires a holistic characterization of subsurface conditions. Delineation of contaminant distribution in all phases (i.e., aqueous, non-aqueous liquid, sorbed, and gas), as well as associated biogeochemical processes in a complex heterogeneous subsurface, is central to selecting effective remedies. Arguably, a factor contributing to the lack of success of managing contaminated sites effectively has been the limitations of site characterization methods that rely on monitoring wells and grab sediment samples. The overarching objective of this research is to advance a set of third-generation (3G) site characterization methods to overcome shortcomings of current site characterization techniques. 3G methods include 1) cryogenic core collection (C3) from unconsolidated geological subsurface to improve recovery of sediments and preserving key attributes, 2) high-throughput analysis (HTA) of frozen core in the laboratory to provide high-resolution, depth discrete data of subsurface conditions and processes, 3) resolution of non-aqueous phase liquid (NAPL) distribution within the porous media using a nuclear magnetic resonance (NMR) method, and 4) application of a complex resistivity method to track NAPL depletion in shallow geological formation over time. A series of controlled experiments were conducted to develop the C 3 tools and methods. The critical aspects of C3 are downhole circulation of liquid nitrogen via a cooling system, the strategic use of thermal insulation to focus cooling into the core, and the use of back pressure to optimize cooling. The C3 methods were applied at two contaminated sites: 1) F.E. Warren (FEW) Air Force Base near Cheyenne, WY and 2) a former refinery in the western U.S. The results indicated that the rate of core collection using the C3 methods is on the order of 30 foot/day. The C3 methods also improve core recovery and limits potential biases associated with flowing sands

  2. Evaluation of reactivity shutdown margin for nuclear fuel reload optimization

    International Nuclear Information System (INIS)

    Wong, Hing-Ip; Maldonado, G.I.

    1995-01-01

    The FORMOSA-P code is a nuclear fuel management optimization package that combines simulated annealing (SA) and nodal generalized perturbation theory (GPT). Recent studies at Electricite de France (EdF-Clamart) have produced good results for power-peaking minimizations under multiple limiting control rod configurations. However, since the reactivity shutdown margin is not explicitly treated as an objective or constraint function, then any optimal loading patterns (LPs) are not guaranteed to yield an adequate shutdown margin (SDM). This study describes the implementation of the SDM calculation within a FORMOSA-P optimization. Maintaining all additional computational requirements to a minimum was a key consideration

  3. Evaluation of reactivity shutdown margin for nuclear fuel reload optimization

    International Nuclear Information System (INIS)

    Engrand, P.; Wong, H. I.; Maldonado, G.I.

    1996-01-01

    The FORMOSA-P code is a nuclear fuel management optimization package which combines simulated annealing (SA) and nodal generalized perturbation theory (GPT). Recent studies at Electricite de France have produced good results for power peaking minimizations under multiple limiting control rod configurations. However, since the reactivity shutdown margin is not explicitly treated as an objective or constraint function, then any optimal loading patterns (LPs) are not guaranteed to yield an adequate shutdown margin (SDM). This study describes the implementation of the SDM calculation within a FORMOSA-P optimization. Maintaining all additional computational requirements to a minimum was a key consideration. (authors). 4 refs., 2 figs

  4. Optimizing Nuclear Reaction Analysis (NRA) using Bayesian Experimental Design

    International Nuclear Information System (INIS)

    Toussaint, Udo von; Schwarz-Selinger, Thomas; Gori, Silvio

    2008-01-01

    Nuclear Reaction Analysis with 3 He holds the promise to measure Deuterium depth profiles up to large depths. However, the extraction of the depth profile from the measured data is an ill-posed inversion problem. Here we demonstrate how Bayesian Experimental Design can be used to optimize the number of measurements as well as the measurement energies to maximize the information gain. Comparison of the inversion properties of the optimized design with standard settings reveals huge possible gains. Application of the posterior sampling method allows to optimize the experimental settings interactively during the measurement process.

  5. Optimization in the nuclear fuel cycle II: Surface contamination

    International Nuclear Information System (INIS)

    Pereira, W.S.; Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Fernandes, T.S.; Mello, C.R.; Kelecom, A.

    2017-01-01

    Optimization is one of the bases of radioprotection and aims to move doses away from the dose limit that is the borderline of acceptable radiological risk. This work aims to use the monitoring of surface contamination as a tool of the optimization process. 53 surface contamination points were analyzed at a nuclear fuel cycle facility. Three sampling points were identified with monthly mean values of contamination higher than 1 Bq ∙ cm -2 , points 28, 42 and 47. These points were indicated for the beginning of the optimization process

  6. Improvements to the sodium supply system of a nuclear reactor core

    International Nuclear Information System (INIS)

    Chevallier, Rene; Marchais, Christian.

    1981-01-01

    This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

  7. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  8. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    International Nuclear Information System (INIS)

    Hedayat, Afshin; Davilu, Hadi; Barfrosh, Ahmad Abdollahzadeh; Sepanloo, Kamran

    2009-01-01

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  9. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aut.ac.i [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of); Davilu, Hadi [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Barfrosh, Ahmad Abdollahzadeh [Department of Computer Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Sepanloo, Kamran [Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of)

    2009-12-15

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  10. Optimal design of nuclear mechanical dampers with analytical hierarchy process

    International Nuclear Information System (INIS)

    Zou Yuehua; Wen Bo; Xu Hongxiang; Qin Yonglie

    2000-01-01

    An optimal design with analytical hierarchy process on nuclear mechanical dampers manufactured by authors' university was described. By using fuzzy judgement matrix the coincidence was automatically satisfied without the need of coincidence test. The results obtained by this method have been put into the production practices

  11. Optimization of radiological protection in Spanish nuclear power plants

    International Nuclear Information System (INIS)

    O'Donnell, P.; Amor, I.; Butragueno, J.L.

    1997-01-01

    Optimizing the radiological protection of occupationally exposed nuclear power plant workers has become one further item in what is called the safety culture. Spanish facilities are implementing programme with this in mind, grounded on a personal motivation policy with the backing of a suitable organizational structure. (Author)

  12. Modeling the optimal management of spent nuclear fuel

    International Nuclear Information System (INIS)

    Nachlas, J.A.; Kurstedt, H.A. Jr.; Swindle, D.W. Jr.; Korcz, K.O.

    1977-01-01

    Recent governmental policy decisions dictate that strategies for managing spent nuclear fuel be developed. Two models are constructed to investigate the optimum residence time and the optimal inventory withdrawal policy for fuel material that presently must be stored. The mutual utility of the models is demonstrated through reference case application

  13. Solving the uncommon nuclear reactor core neutronics problems

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-01-01

    Calculational procedures have been implemented for solving importance and higher harmonic neutronics problems. Solutions are obtained routinely to support analysis of reactor core performance, treating up to three space coordinates with the multigroup diffusion theory approximation to neutron transport. The techniques used and some of the calculational difficulties are discussed

  14. Analysis of the Ford Nuclear Reactor LEU core

    Energy Technology Data Exchange (ETDEWEB)

    Rathkopf, J A; Drumm, C R; Martin, W R; Lee, J C [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1983-09-01

    This paper has summarized the current status of the effort to analyze the FNR HEU/LEU cores and to compare the calculated results with measurements. In general, calculated predictions of experimental results are quite good, especially for global parameters such as reactivity, as seen in the single HEU/LEU element substitution experiment and the LEU full core critical loading. Shim rod worths are predicted well for two of the rods but too high for a third rod possibly due to inaccurate thermal flux distribution calculation. The calculated thermal flux maps show excellent agreement with experiment throughout the FNR core. In the heavy water tank, however, experimental values for the thermal flux obtained by different methods are inconsistent among themselves as well as with the calculated finding. Work is under.way to use our computational tools to correct the discrepancies between the various measurement techniques and to improve the computational results for flux distribution and the rod worth experiment. Although uncertainties exist in our analysis, as evidenced by the discrepancies mentioned above, we consider our present calculational package to be a useful, reasonably accurate, and efficient system for performing analyses of MTR LEU/HEU core configurations.

  15. Piecewise linear approximation: application to control rod step counting in a nuclear reactor core and image contours characterization

    International Nuclear Information System (INIS)

    Kaoutar, M.

    1986-09-01

    After a survey of main algorithms for piecewise linear approximation, a new method is suggested. It consists of two stages: a sequential detection stage and an optimization stage, which derives from general dynamic clustering principle. It is applied to control rod step counting in a nuclear reactor core and images contours characterization. Another version of our method is presented. Its originality cames from the variability of the line segments number during iterations. A comparative study is made by comparing the results of the proposed method with of another methods already existing thereby it attests the efficiency and reliability of our method [fr

  16. Nucleoporins as components of the nuclear pore complex core structure and Tpr as the architectural element of the nuclear basket.

    Science.gov (United States)

    Krull, Sandra; Thyberg, Johan; Björkroth, Birgitta; Rackwitz, Hans-Richard; Cordes, Volker C

    2004-09-01

    The vertebrate nuclear pore complex (NPC) is a macromolecular assembly of protein subcomplexes forming a structure of eightfold radial symmetry. The NPC core consists of globular subunits sandwiched between two coaxial ring-like structures of which the ring facing the nuclear interior is capped by a fibrous structure called the nuclear basket. By postembedding immunoelectron microscopy, we have mapped the positions of several human NPC proteins relative to the NPC core and its associated basket, including Nup93, Nup96, Nup98, Nup107, Nup153, Nup205, and the coiled coil-dominated 267-kDa protein Tpr. To further assess their contributions to NPC and basket architecture, the genes encoding Nup93, Nup96, Nup107, and Nup205 were posttranscriptionally silenced by RNA interference (RNAi) in HeLa cells, complementing recent RNAi experiments on Nup153 and Tpr. We show that Nup96 and Nup107 are core elements of the NPC proper that are essential for NPC assembly and docking of Nup153 and Tpr to the NPC. Nup93 and Nup205 are other NPC core elements that are important for long-term maintenance of NPCs but initially dispensable for the anchoring of Nup153 and Tpr. Immunogold-labeling for Nup98 also results in preferential labeling of NPC core regions, whereas Nup153 is shown to bind via its amino-terminal domain to the nuclear coaxial ring linking the NPC core structures and Tpr. The position of Tpr in turn is shown to coincide with that of the nuclear basket, with different Tpr protein domains corresponding to distinct basket segments. We propose a model in which Tpr constitutes the central architectural element that forms the scaffold of the nuclear basket.

  17. Optimization of multi-phase compressible lattice Boltzmann codes on massively parallel multi-core systems

    NARCIS (Netherlands)

    Biferale, L.; Mantovani, F.; Pivanti, M.; Pozzati, F.; Sbragaglia, M.; Schifano, S.F.; Toschi, F.; Tripiccione, R.

    2011-01-01

    We develop a Lattice Boltzmann code for computational fluid-dynamics and optimize it for massively parallel systems based on multi-core processors. Our code describes 2D multi-phase compressible flows. We analyze the performance bottlenecks that we find as we gradually expose a larger fraction of

  18. Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

    International Nuclear Information System (INIS)

    Darilek, P.

    1991-01-01

    The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

  19. Optimization of heat supply systems employing nuclear power plants

    International Nuclear Information System (INIS)

    Urbanek, J.

    1988-01-01

    Decision making on the further development of heat supply systems requires optimization of the parameters. In particular, meeting the demands of peak load ranges is of importance. The heat supply coefficient α and the annual utilization of peak load equipment τ FS have been chosen as the characteristic quantities to describe them. The heat price at the consumer, C V , offers as the optimization criterion. The transport distance, temperature spread of the heating water, and different curves of annual variation of heat consumption on heat supply coefficient and heat price at the consumer. A comparison between heat supply by nuclear power plants and nuclear heating stations verifies the advantage of combined heat and power generation even with longer heat transport distances as compared with local heat supply by nuclear district heating stations based on the criterion of minimum employment of peak load boilers. (author)

  20. Concept for Multi-cycle Nuclear Fuel Optimization Based On Parallel Simulated Annealing With Mixing of States

    International Nuclear Information System (INIS)

    Kropaczek, David J.

    2008-01-01

    A new concept for performing nuclear fuel optimization over a multi-cycle planning horizon is presented. The method provides for an implicit coupling between traditionally separate in-core and out-of-core fuel management decisions including determination of: fresh fuel batch size, enrichment and bundle design; exposed fuel reuse; and core loading pattern. The algorithm uses simulated annealing optimization, modified with a technique called mixing of states that allows for deployment in a scalable parallel environment. Analysis of algorithm performance for a transition cycle design (i.e. a PWR 6 month cycle length extension) demonstrates the feasibility of the approach as a production tool for fuel procurement and multi-cycle core design. (authors)

  1. A new optimization method based on cellular automata for VVER-1000 nuclear reactor loading pattern

    International Nuclear Information System (INIS)

    Fadaei, Amir Hosein; Setayeshi, Saeed

    2009-01-01

    This paper presents a new and innovative optimization technique, which uses cellular automata for solving multi-objective optimization problems. Due to its ability in simulating the local information while taking neighboring effects into account, the cellular automata technique is a powerful tool for optimization. The fuel-loading pattern in nuclear reactor cores is a major optimization problem. Due to the immensity of the search space in fuel management optimization problems, finding the optimum solution requires a huge amount of calculations in the classical method. The cellular automata models, based on local information, can reduce the computations significantly. In this study, reducing the power peaking factor, while increasing the initial excess reactivity inside the reactor core of VVER-1000, which are two apparently contradictory objectives, are considered as the objective functions. The result is an optimum configuration, which is in agreement with the pattern proposed by the designer. In order to gain confidence in the reliability of this method, the aforementioned problem was also solved using neural network and simulated annealing, and the results and procedures were compared.

  2. Optimal management of fuel in nuclear reactors with slightly enriched uranium and heavy water

    International Nuclear Information System (INIS)

    Serghiuta, D.

    1994-01-01

    This Ph.D. thesis presents the general principles guiding the optimal management of the fuel in CANDU type reactors with slightly enriched uranium. A method is devised which is based on the specific physical characteristics of this type of reactors and makes use of the multipurpose mathematical programming satisfying economical and nuclear safety requirements. The main goal of this work was the establishing of a refueling optimal methodology at equilibrium maintaining the reactor critical during operation. It also minimizes the fuel cycle cost through minimization of the utilized fissile material and at the same time by maximizing the reactor duty time through an optimal chain of refilling operations. This work can be considered as a contribution to a future project of CANDU type reactor core based on slightly enriched uranium. 74 Figs., 9 Tabs., 62 Refs

  3. Optimization of the injection molding process for development of high performance calcium oxide -based ceramic cores

    Science.gov (United States)

    Zhou, P. P.; Wu, G. Q.; Tao, Y.; Cheng, X.; Zhao, J. Q.; Nan, H.

    2018-02-01

    The binder composition used for ceramic injection molding plays a crucial role on the final properties of sintered ceramic and to avoid defects on green parts. In this study, the effects of binder compositions on the rheological, microstructures and the mechanical properties of CaO based ceramic cores were investigated. It was found that the optimized formulation for dispersant, solid loading was 1.5 wt% and 84 wt%, respectively. The microstructures, such as porosity, pore size distribution and grain boundary density were closely related to the plasticizer contents. The decrease of plasticizer contents can enhance the strength of the ceramic cores but with decreased shrinkage. Meanwhile, the creep resistance of ceramic cores was enhanced by decreasing of plasticizer contents. The flexural strength of the core was found to decrease with the increase of the porosity, the improvement of creep resistance is closely related to the decrease of porosity and grain boundary density.

  4. Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

    2007-07-01

    Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

  5. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

  6. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  7. The effects of core zoning on optimization of design analysis of molten salt reactor

    International Nuclear Information System (INIS)

    Guo, Zhangpeng; Wang, Chenglong; Zhang, Dalin; Chaudri, Khurrum Saleem; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2013-01-01

    Highlights: • 1/8 of core is simulated by MCNP and thermal-hydraulic code simultaneously. • Effects of core zoning are studied by dividing the core into two regions. • Both the neutronics and thermal-hydraulic behavior are investigated. • The flat flux distribution is achieved in the optimization analysis. • The flat flux can lead to worse thermal-hydraulic behavior occasionally. - Abstract: The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly

  8. Core safety of Indian nuclear power plants (NPPs) under extreme ...

    Indian Academy of Sciences (India)

    Of course, the accidents led to release of radioactivity due to probable melt down of reactor ... advances in technology and a better understanding of the nuclear power, the ..... PHWR system offers certain intrinsic advantages (Narora Atomic Power ...... system, Non Active Process Water System (NAPWS), Service Water Sys-.

  9. Pinning down nuclear. To the core of the matter

    International Nuclear Information System (INIS)

    Boeck, Helmut; Gerstmayr, Michael; Radde, Eileen

    2014-01-01

    The nuclear disaster in Fukushima shocked the world tremendously. The call to pull out of nuclear energy is getting louder - and more often than not by politicians trying to lure the favour of voters. Through the media there are half-truths and false information floating about the global consequences of the disaster and sensational prognoses for the future, all of which are in turn unsettling for the general public. Are the opposers to nuclear energy playing with the fear of the public or is the threat real? This book tells, in a captivating manner - authenticated with examples and incidents not known by many - what the threat for the area actually looks like. They confront the level of truth in the frightening scenarios and inform about the situation in case of emergency. Furthermore, they examine factors that preceded the disaster and broach the subject of the incredible hunger for energy, which dominates the world and continues to drive the commercial use of nuclear energy. Also the ghost of Chernobyl and its aftermath, which has been dismissed from our minds, is re-examined based on current knowledge. The book impresses with insider know-how, latest detailed knowledge, amazing facts and an entertaining narrative style.

  10. Pinning down nuclear. To the core of the matter

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, Helmut; Gerstmayr, Michael [Technische Univ., Vienna (Austria); International Atomic Energy Agency, Vienna (Austria); Radde, Eileen [Nuclear Engineering Seibersdorf GmbH (Austria); International Atomic Energy Agency, Vienna (Austria)

    2014-07-01

    The nuclear disaster in Fukushima shocked the world tremendously. The call to pull out of nuclear energy is getting louder - and more often than not by politicians trying to lure the favour of voters. Through the media there are half-truths and false information floating about the global consequences of the disaster and sensational prognoses for the future, all of which are in turn unsettling for the general public. Are the opposers to nuclear energy playing with the fear of the public or is the threat real? This book tells, in a captivating manner - authenticated with examples and incidents not known by many - what the threat for the area actually looks like. They confront the level of truth in the frightening scenarios and inform about the situation in case of emergency. Furthermore, they examine factors that preceded the disaster and broach the subject of the incredible hunger for energy, which dominates the world and continues to drive the commercial use of nuclear energy. Also the ghost of Chernobyl and its aftermath, which has been dismissed from our minds, is re-examined based on current knowledge. The book impresses with insider know-how, latest detailed knowledge, amazing facts and an entertaining narrative style.

  11. Heuristic rules embedded genetic algorithm for in-core fuel management optimization

    Science.gov (United States)

    Alim, Fatih

    The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code

  12. A fuzzy controller design for nuclear research reactors using the particle swarm optimization algorithm

    International Nuclear Information System (INIS)

    Coban, Ramazan

    2011-01-01

    Research highlights: → A closed-loop fuzzy logic controller based on the particle swarm optimization algorithm was proposed for controlling the power level of nuclear research reactors. → The proposed control system was tested for various initial and desired power levels, and it could control the reactor successfully for most situations. → The proposed controller is robust against the disturbances. - Abstract: In this paper, a closed-loop fuzzy logic controller based on the particle swarm optimization algorithm is proposed for controlling the power level of nuclear research reactors. The principle of the fuzzy logic controller is based on the rules constructed from numerical experiments made by means of a computer code for the core dynamics calculation and from human operator's experience and knowledge. In addition to these intuitive and experimental design efforts, consequent parts of the fuzzy rules are optimally (or near optimally) determined using the particle swarm optimization algorithm. The contribution of the proposed algorithm to a reactor control system is investigated in details. The performance of the controller is also tested with numerical simulations in numerous operating conditions from various initial power levels to desired power levels, as well as under disturbance. It is shown that the proposed control system performs satisfactorily under almost all operating conditions, even in the case of very small initial power levels.

  13. Management of nuclear PRs activity with optimal conditions

    International Nuclear Information System (INIS)

    Ohnishi, Teruaki

    1997-01-01

    A methodology is proposed to derive optimal conditions for the activity of nuclear public relations (PRs). With the use of data-bases available at present, expressions were derived which connect the budget allocated for the PRs activity with the intensity of stimulus for four types of activity of the advertisement in the press, the exclusive publicity, the pamphlet and the advertisement on television. Optimal conditions for the activity were determined by introducing a model describing a relation between the intensity of stimulus and the extent of the change of public's attitude to nuclear energy, namely the effect of PRs activity, and also by giving the optimal ratio of allocation of the budget among the four types of activity as a function of cost versus effectiveness of each type. Those optimal conditions, being for the ratio of allocation of the budget, the execution time and the intensity of each type of activity at that time, vary depending on the number of household in a target region, the target class of demography, the duration time of activity, and the amount of budget for the activity. It becomes clear from numerical calculation that the optimal conditions and the effect of activity show quite strong non-linearity with respect to the variation of those variables, and that the effect of PRs activity averaged over all public in the target region becomes to be maximum, in Japan, when the activity is executed with the optimal conditions determined for the target class of middle- and advanced-aged women. The management of nuclear PRs activity becomes possible by introducing such a method of fixation of optimal conditions for the activity as described here. (author)

  14. Pressurized water reactor in-core nuclear fuel management by tabu search

    International Nuclear Information System (INIS)

    Hill, Natasha J.; Parks, Geoffrey T.

    2015-01-01

    Highlights: • We develop a tabu search implementation for PWR reload core design. • We conduct computational experiments to find optimal parameter values. • We test the performance of the algorithm on two representative PWR geometries. • We compare this performance with that given by established optimization methods. • Our tabu search implementation outperforms these methods in all cases. - Abstract: Optimization of the arrangement of fuel assemblies and burnable poisons when reloading pressurized water reactors has, in the past, been performed with many different algorithms in an attempt to make reactors more economic and fuel efficient. The use of the tabu search algorithm in tackling reload core design problems is investigated further here after limited, but promising, previous investigations. The performance of the tabu search implementation developed was compared with established genetic algorithm and simulated annealing optimization routines. Tabu search outperformed these existing programs for a number of different objective functions on two different representative core geometries

  15. A computational technique to identify the optimal stiffness matrix for a discrete nuclear fuel assembly model

    International Nuclear Information System (INIS)

    Park, Nam-Gyu; Kim, Kyoung-Joo; Kim, Kyoung-Hong; Suh, Jung-Min

    2013-01-01

    Highlights: ► An identification method of the optimal stiffness matrix for a fuel assembly structure is discussed. ► The least squares optimization method is introduced, and a closed form solution of the problem is derived. ► The method can be expanded to the system with the limited number of modes. ► Identification error due to the perturbed mode shape matrix is analyzed. ► Verification examples show that the proposed procedure leads to a reliable solution. -- Abstract: A reactor core structural model which is used to evaluate the structural integrity of the core contains nuclear fuel assembly models. Since the reactor core consists of many nuclear fuel assemblies, the use of a refined fuel assembly model leads to a considerable amount of computing time for performing nonlinear analyses such as the prediction of seismic induced vibration behaviors. The computational time could be reduced by replacing the detailed fuel assembly model with a simplified model that has fewer degrees of freedom, but the dynamic characteristics of the detailed model must be maintained in the simplified model. Such a model based on an optimal design method is proposed in this paper. That is, when a mass matrix and a mode shape matrix are given, the optimal stiffness matrix of a discrete fuel assembly model can be estimated by applying the least squares minimization method. The verification of the method is completed by comparing test results and simulation results. This paper shows that the simplified model's dynamic behaviors are quite similar to experimental results and that the suggested method is suitable for identifying reliable mathematical model for fuel assemblies

  16. A new approach to nuclear reactor design optimization using genetic algorithms and regression analysis

    International Nuclear Information System (INIS)

    Kumar, Akansha; Tsvetkov, Pavel V.

    2015-01-01

    Highlights: • This paper presents a new method useful for the optimization of complex dynamic systems. • The method uses the strengths of; genetic algorithms (GA), and regression splines. • The method is applied to the design of a gas cooled fast breeder reactor design. • Tools like Java, R, and codes like MCNP, Matlab are used in this research. - Abstract: A module based optimization method using genetic algorithms (GA), and multivariate regression analysis has been developed to optimize a set of parameters in the design of a nuclear reactor. GA simulates natural evolution to perform optimization, and is widely used in recent times by the scientific community. The GA fits a population of random solutions to the optimized solution of a specific problem. In this work, we have developed a genetic algorithm to determine the values for a set of nuclear reactor parameters to design a gas cooled fast breeder reactor core including a basis thermal–hydraulics analysis, and energy transfer. Multivariate regression is implemented using regression splines (RS). Reactor designs are usually complex and a simulation needs a significantly large amount of time to execute, hence the implementation of GA or any other global optimization techniques is not feasible, therefore we present a new method of using RS in conjunction with GA. Due to using RS, we do not necessarily need to run the neutronics simulation for all the inputs generated from the GA module rather, run the simulations for a predefined set of inputs, build a multivariate regression fit to the input and the output parameters, and then use this fit to predict the output parameters for the inputs generated by GA. The reactor parameters are given by the, radius of a fuel pin cell, isotopic enrichment of the fissile material in the fuel, mass flow rate of the coolant, and temperature of the coolant at the core inlet. And, the optimization objectives for the reactor core are, high breeding of U-233 and Pu-239 in

  17. Nuclear design and analysis report for KALIMER breakeven core conceptual design

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Song, Hoon; Lee, Ki Bog; Chang, Jin Wook; Hong, Ser Gi; Kim, Young Gyun; Kim, Yeong Il

    2002-04-01

    During the phase 2 of LMR design technology development project, the breakeven core configuration was developed with the aim of the KALIMER self-sustaining with regard to the fissile material. The excess fissile material production is limited only to the extent of its own requirement for sustaining its planned power operation. The average breeding ratio is estimated to be 1.05 for the equilibrium core and the fissile plutonium gain per cycle is 13.9 kg. The nuclear performance characteristics as well as the reactivity coefficients have been analyzed so that the design evaluation in other activity areas can be made. In order to find out a realistic heavy metal flow evolution and investigate cycle-dependent nuclear performance parameter behaviors, the startup and transition cycle loading strategies are developed, followed by the startup core physics analysis. Driver fuel and blankets are assumed to be shuffled at the time of each reload. The startup core physics analysis has shown that the burnup reactivity swing, effective delayed neutron fraction, conversion ratio and peak linear heat generation rate at the startup core lead to an extreme of bounding physics data for safety analysis. As an outcome of this study, a whole spectrum of reactor life is first analyzed in detail for the KALIMER core. It is experienced that the startup core analysis deserves more attention than the current design practice, before the core configuration is finalized based on the equilibrium cycle analysis alone.

  18. Optimization and photomodification of extremely broadband optical response of plasmonic core-shell obscurants.

    Science.gov (United States)

    de Silva, Vashista C; Nyga, Piotr; Drachev, Vladimir P

    2016-12-15

    Plasmonic resonances of the metallic shells depend on their nanostructure and geometry of the core, which can be optimized for the broadband extinction normalized by mass. The fractal nanostructures can provide a broadband extinction. It allows as well for a laser photoburning of holes in the extinction spectra and consequently windows of transparency in a controlled manner. The studied core-shell microparticles synthesized using colloidal chemistry consist of gold fractal nanostructures grown on precipitated calcium carbonate (PCC) microparticles or silica (SiO 2 ) microspheres. The optimization includes different core sizes and shapes, and shell nanostructures. It shows that the rich surface of the PCC flakes is the best core for the fractal shells providing the highest mass normalized extinction over the extremely broad spectral range. The mass normalized extinction cross section up to 3m 2 /g has been demonstrated in the broad spectral range from the visible to mid-infrared. Essentially, the broadband response is a characteristic feature of each core-shell microparticle in contrast to a combination of several structures resonant at different wavelengths, for example nanorods with different aspect ratios. The photomodification at an IR wavelength makes the window of transparency at the longer wavelength side. Copyright © 2016 Elsevier Inc. All rights reserved.

  19. Optimization of core reload design for low leakage fuel management in pressurized water reactors

    International Nuclear Information System (INIS)

    Kim, Y.J.

    1986-01-01

    A new method was developed to optimize pressurized water reactor core reload design for low leakage fuel management, a strategy recently adopted by most utilities to extend cycle length and mitigate pressurized thermal shock concerns. The method consists of a two-stage optimization process which provides the maximum cycle length for a given fresh fuel loading subject to power peaking constraints. In the first stage, a best fuel arrangement is determined at the end of cycle in the absence of burnable poisons. A direct search method is employed in conjunction with a constant power, Haling depletion. In the second stage, the core control poison requirements are determined using a linear programming technique. The solution provides the fresh fuel burnable poison loading required to meet core power peaking constraints. An accurate method of explicitly modeling burnable absorbers was developed for this purpose. The design method developed here was implemented in a currently recognized fuel licensing code, SIMULATE, that was adapted to the CYBER-205 computer. This methodology was applied to core reload design of cycles 9 and 10 for the Commonwealth Edison Zion, Unit-1 Reactor. The results showed that the optimum loading pattern for cycle 9 yielded almost a 9% increase in the cycle length while reducing core vessel fluence by 30% compared with the reference design used by Commonwealth Edison

  20. Notes on nuclear reactor core analysis code: CITATION

    International Nuclear Information System (INIS)

    Cepraga, D.G.

    1980-01-01

    The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)

  1. Nuclear reactor ex-core startup neutron detector

    International Nuclear Information System (INIS)

    Wyvill, J.R.

    1980-01-01

    A sensitive ex-core neutron detector is needed to monitor the power level of reactors during startup. The neutron detector of this invention has a photomultiplier with window resistant to radiation darkening at the input end and an electrical connector at the output end. The photomultiplier receives light signals from a neutron-responsive scintillator medium, typically a cerium-doped lithium silicate glass, that responds to neutrons after they have been thermalized by a silicone resin moderator. Enclosing and shielding the photmultiplier, the scintillator medium and the moderator is a combined lead and borated silicone resin housing

  2. A discrete optimization method for nuclear fuel management

    International Nuclear Information System (INIS)

    Argaud, J.P.

    1993-04-01

    Nuclear loading pattern elaboration can be seen as a combinational optimization problem of tremendous size and with non-linear cost-functions, and search are always numerically expensive. After a brief introduction of the main aspects of nuclear fuel management, this paper presents a new idea to treat the combinational problem by using informations included in the gradient of a cost function. The method is to choose, by direct observation of the gradient, the more interesting changes in fuel loading patterns. An example is then developed to illustrate an operating mode of the method, and finally, connections with simulated annealing and genetic algorithms are described as an attempt to improve search processes

  3. Two-dimensional core calculation research for fuel management optimization based on CPACT code

    International Nuclear Information System (INIS)

    Chen Xiaosong; Peng Lianghui; Gang Zhi

    2013-01-01

    Fuel management optimization process requires rapid assessment for the core layout program, and the commonly used methods include two-dimensional diffusion nodal method, perturbation method, neural network method and etc. A two-dimensional loading patterns evaluation code was developed based on the three-dimensional LWR diffusion calculation program CPACT. Axial buckling introduced to simulate the axial leakage was searched in sub-burnup sections to correct the two-dimensional core diffusion calculation results. Meanwhile, in order to get better accuracy, the weight equivalent volume method of the control rod assembly cross-section was improved. (authors)

  4. Optimization of High-Resolution Continuous Flow Analysis for Transient Climate Signals in Ice Cores

    DEFF Research Database (Denmark)

    Bigler, Matthias; Svensson, Anders; Kettner, Ernesto

    2011-01-01

    Over the past two decades, continuous flow analysis (CFA) systems have been refined and widely used to measure aerosol constituents in polar and alpine ice cores in very high-depth resolution. Here we present a newly designed system consisting of sodium, ammonium, dust particles, and electrolytic...... meltwater conductivity detection modules. The system is optimized for high- resolution determination of transient signals in thin layers of deep polar ice cores. Based on standard measurements and by comparing sections of early Holocene and glacial ice from Greenland, we find that the new system features...

  5. Genetic algorithm for the optimization of the loading pattern for reactor core fuel management

    International Nuclear Information System (INIS)

    Zhou Sheng; Hu Yongming; zheng Wenxiang

    2000-01-01

    The paper discusses the application of a genetic algorithm to the optimization of the loading pattern for in-core fuel management with the NP characteristics. The algorithm develops a matrix model for the fuel assembly loading pattern. The burnable poisons matrix was assigned randomly considering the distributed nature of the poisons. A method based on the traveling salesman problem was used to solve the problem. A integrated code for in-core fuel management was formed by combining this code with a reactor physics code

  6. Optimization of the coherence function estimation for multi-core central processing unit

    Science.gov (United States)

    Cheremnov, A. G.; Faerman, V. A.; Avramchuk, V. S.

    2017-02-01

    The paper considers use of parallel processing on multi-core central processing unit for optimization of the coherence function evaluation arising in digital signal processing. Coherence function along with other methods of spectral analysis is commonly used for vibration diagnosis of rotating machinery and its particular nodes. An algorithm is given for the function evaluation for signals represented with digital samples. The algorithm is analyzed for its software implementation and computational problems. Optimization measures are described, including algorithmic, architecture and compiler optimization, their results are assessed for multi-core processors from different manufacturers. Thus, speeding-up of the parallel execution with respect to sequential execution was studied and results are presented for Intel Core i7-4720HQ и AMD FX-9590 processors. The results show comparatively high efficiency of the optimization measures taken. In particular, acceleration indicators and average CPU utilization have been significantly improved, showing high degree of parallelism of the constructed calculating functions. The developed software underwent state registration and will be used as a part of a software and hardware solution for rotating machinery fault diagnosis and pipeline leak location with acoustic correlation method.

  7. Nuclear determination of saturation profiles in core plugs

    International Nuclear Information System (INIS)

    Sletsgaard, J.; Oelgaard, P.L.

    1997-01-01

    A method to determine liquid saturations in core plugs during flooding is of importance when the relative permeability and capillary pressure function are to be determined. This part of the EFP-95 project uses transmission of γ-radiation to determine these saturations. In γ-transmission measurements, the electron density of the given substance is measured. This is an advantage as compared to methods that use electric conductivity, since neither oil nor gas conducts electricity. At the moment a single 137 Cs-source is used, but a theoretical investigation of whether it is possible to determine three saturations, using two radioactive sources with different γ-energies, has been performed. Measurements were made on three core plugs. To make sure that the measurements could be reproduced, all the plugs had a point of reference, i.e. a mark so that it was possible to place the plug same way every time. Two computer programs for calculation of saturation and porosity and the experimental setup are listed. (EG)

  8. The density jump at the inner core boundary using underground nuclear explosion records

    International Nuclear Information System (INIS)

    Krasnoshchekov, D.N.; Ovchinnikov, V.M.

    2001-01-01

    This paper presents the estimation of the minimum jump value using experimental wave forms reflected from the boundary between the Earth core and mantle (PcP) and the one between the inner and outer core (PKiKP) at a distance of 6 deg. Digital seismic records of underground nuclear tests conducted at the Semipalatinsk test site in 70s by Zerenda-Vostochny-Chkalovo seismic array have been used. (author)

  9. Contribution of Anticipated Transients Without Scram (ATWS) to core melt at United States nuclear power plants

    International Nuclear Information System (INIS)

    Giachetti, R.T.

    1989-09-01

    This report looks at WASH-1400 and several other Probabilistic Risk Assessments (PRAs) and Probabilistic Safety Studies (PSSs) to determine the contribution of Anticipated Transients Without Scram (ATWS) events to the total core melt probability at eight nuclear power plants in the United States. After considering each plant individually, the results are compared from plant to plant to see if any generic conclusions regarding ATWS, or core melt in general, can be made. 8 refs., 34 tabs

  10. A novel optimization method, Effective Discrete Firefly Algorithm, for fuel reload design of nuclear reactors

    International Nuclear Information System (INIS)

    Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.

    2015-01-01

    Highlights: • An advanced version of firefly algorithm, EDFA, is proposed for the core pattern optimization problem. • The movement of each firefly toward the best firefly with a dynamic probability is the major improvement of EDFA. • LPO results represent the faster convergence and better performance of EDFA in comparison to CFA and DFA. - Abstract: Inspired by fireflies behavior in nature, a firefly algorithm has been developed for solving optimization problems. In this approach, each firefly movement is based on absorption of the other one. For enhancing the performance of firefly algorithm in the optimization process of nuclear reactor loading pattern optimization (LPO), we introduce a new variant of firefly algorithm, i.e. Effective Discrete Firefly Algorithm (EDFA). In EDFA, a new behavior is the movement of fireflies to current global best position with a dynamic probability, i.e. the movement of each firefly can be determined to be toward the brighter or brightest firefly’s position in any iteration of the algorithm. In this paper, our optimization objectives for the LPO are the maximization of K eff along with the minimization of the power peaking factor (PPF). In order to represent the increase of convergence speed of EDFA, basic firefly algorithms including the continuous firefly algorithm (CFA) and the discrete firefly algorithm (DFA) also have been implemented. Loading pattern optimization results of two well-known problems confirm better performance of EDFA in obtaining nearly optimized fuel arrangements in comparison to CFA and DFA. All in all, we can suggest applying the EDFA to other optimization problems of nuclear engineering field in order to investigate its performance in gaining considered objectives

  11. Optimal improvement of graphs related to nuclear safeguards problems

    International Nuclear Information System (INIS)

    Jacobsen, S.E.

    1977-08-01

    This report develops the methodology for optimally improving graphs related to nuclear safeguards issues. In particular, given a fixed number of dollars, the report provides a method for optimally allocating such dollars over the arcs of a weighted graph (the weights vary as a function of dollars spent on arcs) so as to improve the system effectiveness measure which is the shortest of all shortest paths to several targets. Arc weights can be either clock times or detection probabilities and the algorithm does not explicitly consider all paths to the targets

  12. Optimization of the control of contamination at nuclear power plants

    International Nuclear Information System (INIS)

    Khan, T.A.; Baum, J.W.

    1988-05-01

    A methodology is described for the optimization of the actions taken to control contamination. It deals with many aspects of contamination, such as the monetary value assigned to a unit of radiation dose, the treatment of skin and extremity dose, and the inefficiencies introduced from working in a contaminated environemnt. The optimization method is illustrated with two case studies based on cleanup projects at nuclear power plants. Guidelines for the use of protective apparel, and for monitoring radiation and contamination at various levels of contamination are presented. The report concludes that additional research is required to quantify the effect of a contaminated environment on work efficiencies

  13. The optimization design of nuclear measurement teaching equipment

    International Nuclear Information System (INIS)

    Tang Rulong; Qiu Xiaoping

    2008-01-01

    So far domestic student-oriented experimental nuclear measuring instruments are used only to measure object density, thickness or material level, and in the choice of sources activity is mostly about 10 mCi. this design will proposed a optimization program dealing with domestic situation. It discussed the radioactive sources activity, the structural design of sealed sources, such as the choice of the tested material in order to get a program optimization. The program used 1 mCi activity radioactive sources 137 Cs to reduce the radiation dose, and the measurement function was improved. So that the apparatus can measure density, thickness nad material level. (authors)

  14. Present status of reactor physics in the United States and Japan-III. 2. Nuclear Fuel Management Optimization Capabilities

    International Nuclear Information System (INIS)

    Karve, Atul A.; Keller, Paul M.; Turinsky, Paul J.; Maldonado, G. Ivan

    2001-01-01

    Nuclear fuel management is a very difficult design optimization problem in that decisions ranging from the microscopic level, e.g., pin enrichment, to the macroscopic level, e.g., core flow rate, and spanning time horizons of several reload cycles are strongly coupled. Added to these attributes are the highly constrained design, disjointed decision space, multimodal objective function, mixed integer type decision variables, highly nonlinear objective and constraint functions, and computationally demanding evaluation of the objective and constraint functions. Not surprisingly, after years of research on nuclear fuel management optimization, only limited progress has been made. The traditional approach to partially overcome these difficulties involves constraining the search space via heuristic rules, decomposing the problem into sub-optimization problems, and utilizing simplified core physics models. These approaches have sometimes proven effective, but to claim that the design decisions are global optimum decisions would not be appropriate. Given the increasingly tight constraints and design complexities of nuclear cores, and stronger desire to reduce generating costs, the nuclear fuel management design optimization problem has grown more challenging and important with the passage of time. In this paper, we summarize our research on this design optimization problem. A suite of computer codes that aid in making nuclear fuel management decisions has been developed. From Table I, it is obvious that decomposition of the global optimization problem into suboptimum problems has been employed. All of these computer codes utilize stochastic optimization techniques to search the decision space for determining the family of near-optimum decisions in the sub-optimization problem being solved. A stochastic optimization approach has been selected since it is well suited to address the problems' attributes noted earlier. The drawback of employing a stochastic optimization

  15. Analytic function expansion nodal method for nuclear reactor core design

    International Nuclear Information System (INIS)

    Noh, Hae Man

    1995-02-01

    In most advanced nodal methods the transverse integration is commonly used to reduce the multi-dimensional diffusion equation into equivalent one- dimensional diffusion equations when derving the nodal coupling equations. But the use of the transverse integration results in some limitations. The first limitation is that the transverse leakage term which appears in the transverse integration procedure must be appropriately approximated. The second limitation is that the one-dimensional flux shapes in each spatial direction resulted from the nodal calculation are not accurate enough to be directly used in reconstructing the pinwise flux distributions. Finally the transverse leakage defined for a non-rectangular node such as a hexagonal node or a triangular node is too complicated to be easily handled and may contain non-physical singular terms of step-function and delta-function types. In this thesis, the Analytic Function Expansion Nodal (AFEN) method and its two variations : the Polynomial Expansion Nodal (PEN) method and the hybrid of the AFEN and PEN methods, have been developed to overcome the limitations of the transverse integration procedure. All of the methods solve the multidimensional diffusion equation without the transverse integration. The AFEN method which we believe is the major contribution of this study to the reactor core analysis expands the homogeneous flux distributions within a node in non-separable analytic basis functions satisfying the neutron diffusion equations at any point of the node and expresses the coefficients of the flux expansion in terms of the nodal unknowns which comprise a node-average flux, node-interface fluxes, and corner-point fluxes. Then, the nodal coupling equations composed of the neutron balance equations, the interface current continuity equations, and the corner-point leakage balance equations are solved iteratively to determine all the nodal unknowns. Since the AFEN method does not use the transverse integration in

  16. Impact of nuclear 'pasta' on neutrino transport in collapsing stellar cores

    International Nuclear Information System (INIS)

    Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Takiwaki, Tomoya; Yasuoka, Kenji; Ebisuzaki, Toshikazu

    2007-01-01

    Nuclear 'pasta', nonspherical nuclei in dense matter, is predicted to occur in collapsing supernova cores. We show how pasta phases affect the neutrino transport cross section via weak neutral current using several nuclear models. This is the first calculation of the neutrino opacity of the phases with rod-like and slab-like nuclei taking account of finite temperature effects, which are well described by the quantum molecular dynamics. We also show that pasta phases can occupy 10-20% of the mass of supernova cores in the later stage of the collapse

  17. In-core gamma dosimetry by solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Khan, H.A.

    1980-02-01

    Results are reported of a study undertaken to develop Solid State Nuclear Track Detectors (SSNTD) for the measurement of gamma doses in the megarad region such as those existing in and around a nuclear reactor core. The changes brought about in the track etching parameters and in the ultraviolet and infrared transmittances, have been studied for possible use as gamma dose measuring indices. Effects of various parameters in the core such as neutron flux, beta particles, water, temperature, and gamma ray spectrum have been investigated and found to have only small influence on the proposed gamma dose measuring indices

  18. Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

    International Nuclear Information System (INIS)

    Castiblanco, L.A.; Sarta, J.A.

    1997-01-01

    The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

  19. The PBIL algorithm applied to a nuclear reactor design optimization

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Marcelo D.; Medeiros, Jose A.C.C.; Lima, Alan M.M. de; Schirru, Roberto [Instituto Alberto Luiz Coimbra de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ-RJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear. Lab. de Monitoracao de Processos]. E-mails: marcelo@lmp.ufrj.br; canedo@lmp.ufrj.br; alan@lmp.ufrj.br; schirru@lmp.ufrj.br

    2007-07-01

    The Population-Based Incremental Learning (PBIL) algorithm is a method that combines the mechanism of genetic algorithm with the simple competitive learning, creating an important tool to be used in the optimization of numeric functions and combinatory problems. PBIL works with a set of solutions to the problems, called population, whose objective is create a probability vector, containing real values in each position, that when used in a decoding procedure gives subjects that present the best solutions for the function to be optimized. In this work a new form of learning for algorithm PBIL is developed, having aimed at to reduce the necessary time for the optimization process. This new algorithm will be used in the nuclear reactor design optimization. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment zone reactor, considering some restrictions. In this optimization is used the computational code HAMMER, and the results compared with other methods of optimization by artificial intelligence. (author)

  20. The PBIL algorithm applied to a nuclear reactor design optimization

    International Nuclear Information System (INIS)

    Machado, Marcelo D.; Medeiros, Jose A.C.C.; Lima, Alan M.M. de; Schirru, Roberto

    2007-01-01

    The Population-Based Incremental Learning (PBIL) algorithm is a method that combines the mechanism of genetic algorithm with the simple competitive learning, creating an important tool to be used in the optimization of numeric functions and combinatory problems. PBIL works with a set of solutions to the problems, called population, whose objective is create a probability vector, containing real values in each position, that when used in a decoding procedure gives subjects that present the best solutions for the function to be optimized. In this work a new form of learning for algorithm PBIL is developed, having aimed at to reduce the necessary time for the optimization process. This new algorithm will be used in the nuclear reactor design optimization. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a 3-enrichment zone reactor, considering some restrictions. In this optimization is used the computational code HAMMER, and the results compared with other methods of optimization by artificial intelligence. (author)

  1. Pattern recognition of subcooled boiling in core of nuclear reactors

    International Nuclear Information System (INIS)

    Inyushev, V.V.; Sharaevskij, I.G.

    2003-01-01

    The noise signals at the outlet of main nuclear power plant technological parameter gauges (neutron flux, dynamic pressure etc.) contain important information on technical state of the equipment. In the work efficient algorithms of random process identification that after respective spectral transformation are considered as multidimensional random vectors were developed. Automated classification of these vector in the developed algorithms in realized on the base of probability function, especially of Bayes classifier. Application of classifier is based on construction of multidimensional distribution of probabilities in feature space of corresponding dimension, with the help of which random vectors-realizations of corresponding images that are subjected to automated classification are described

  2. Development of Inventory Optimization System for Operation Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Se-Jin; Park, Jong-Hyuk; Yoo, Sung-Soo; Lee, Sang-Guk [Korea Electric Power Research Institutes, Taejon (Korea, Republic of)

    2006-07-01

    Inventory control of spare parts plays an increasingly important role in operation management. This is why inventory management systems such as manufacturing resources planning(MRP) and enterprise resource planning(ERP) have been added. However, most of these contributions have similar theoretical background. This means the concepts and techniques are mainly based on mathematical assumptions and modeling inventory of spare parts situations. Nuclear utilities in Korea have several problems to manage the optimum level of spare parts though they used MRP System. Because most of items have long lead time and they are imported from United States, Canada, France and so on. We developed the inventory optimization system for Operation Nuclear Plants to resolve these problems. In this paper, we report a data flow process, data load and inventory calculation process. The main contribution of this paper is development of inventory optimization system which can be used in domestic power plants.

  3. Development of Inventory Optimization System for Operation Nuclear Plants

    International Nuclear Information System (INIS)

    Jang, Se-Jin; Park, Jong-Hyuk; Yoo, Sung-Soo; Lee, Sang-Guk

    2006-01-01

    Inventory control of spare parts plays an increasingly important role in operation management. This is why inventory management systems such as manufacturing resources planning(MRP) and enterprise resource planning(ERP) have been added. However, most of these contributions have similar theoretical background. This means the concepts and techniques are mainly based on mathematical assumptions and modeling inventory of spare parts situations. Nuclear utilities in Korea have several problems to manage the optimum level of spare parts though they used MRP System. Because most of items have long lead time and they are imported from United States, Canada, France and so on. We developed the inventory optimization system for Operation Nuclear Plants to resolve these problems. In this paper, we report a data flow process, data load and inventory calculation process. The main contribution of this paper is development of inventory optimization system which can be used in domestic power plants

  4. FFTF reload core nuclear design for increased experimental capability

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Nelson, J.V.; Dobbin, K.D.; Bennett, R.A.

    1976-01-01

    In anticipation of continued growth in the FTR experimental irradiations program, the enrichments for the next batches of reload driver fuel to be manufactured have been increased to provide a substantially enlarged experimental reactivity allowance. The enrichments for these fuel assemblies, termed ''Cores 3 and 4,'' were selected to meet the following objectives and constraints: (1) maintain a reactor power capability of 400 MW (based on an evaluation of driver fuel centerline melting probability at 15 percent overpower); (2) provide a peak neutron flux of nominally 7 x 10 15 n/cm 2 -sec, with a minimum acceptable value of 95 percent of this (i.e., 6.65 x 10 15 n/cm 2 -sec); and (3) provide the maximum experimental reactivity allowance that is consistent with the above constraints

  5. Studies on the inhomogeneous core density of a fluidized bed nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Hagen, T.H.J.J.; Van Dam, H.; Hoogenboom, J.E.; Khotylev, V.A. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.; Harteveld, W.; Mudde, R.F.

    1997-12-31

    Results are reported on the expected time dependent core density profile of a fluidized-bed nuclear fission reactor. Core densities have been measured in a test facility by the gamma-transmission technique. Bubble and particle-cluster sizes, positions, velocities and frequencies could be determined. Neutronic studies have been performed on the influence of core voids on reactivity using Monte-Carlo and neutron-transport codes. Fuel-particle importance has been determined. Point-kinetic parameters have been calculated for linking reactivity perturbations to power fluctuations. (author)

  6. Analysis of a basic core performance for FBR core nuclear design. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    1999-03-01

    The spatial distribution of reaction rates in the ZPPR-13A, having an axially heterogeneous core, has been analyzed. The ZPPR-13A core is treated as a 2-dimensional RZ configuration consisting of a homogeneous core. The analysis is performed by utilizing both probabilistic and deterministic treatments. The probabilistic treatment is performed with the Monte Carlo Code MVP running with continuous energy variable. By comparing the results obtained by both treatments and reviewing the calculation method of effective resonance cross sections, for deterministic treatment, utilized for the reaction rate distributions, it is revealed that the present treatment of effective resonance cross sections is not accurate, since there are observed effects due to dependence on energy group number or energy group width, and on anisotropic scattering. To utilize multi-band method for calculating effective resonance cross sections, widely used by the European researchers, the computer code GROUPIE is installed and the performance of the code is confirmed. Although, in order to improve effective resonance cross sections accuracy, the thermal neutron reactor standard code system SRAC-95 was introduced last year in which the ultra-fine group spectrum calculation module PEACO worked specially under the restriction that number of nuclei having resonance cross section, in any zone, should be less than three, because collision probabilities were obtained by an interpolation method. This year, the module is improved so that these collision probabilities are directly calculated, and by this improvement the highly accurate effective resonance cross sections below the energy of 40.868 keV can be calculated for whole geometrical configurations considered. To extend the application range of the module PEACO, the cross sections of sodium and structure material nuclei are prepared so that they are also represented as ultra-fine group cross sections. By such modifications of cross section library

  7. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  8. Sustaining Nuclear Safety: Upholding the Core Regulatory Values

    International Nuclear Information System (INIS)

    Kumar, S.

    2016-01-01

    Nuclear Energy and management of safety therein, has a somewhat distinct streak in that from its early days it has had the privilege of being shaped and supervised by the eminent scientists and engineers, in fact it owes its very origin to them. This unique engagement has resulted in culmination of the several safety elements like defence-in-depth in the form of multiple safety layers, redundancy, diversity and physical separation of components, protection against single failures as well as common cause failures right at the beginning of designing a nuclear reactor. The fundamental principles followed by regulators across the globe have many similarities such as, creation of an organization which has a conflict-free primary responsibility of safety supervision, laying down the safety criteria and requirements for the respective industry and developing and using various tools and regulatory methodology to ensure adherence to the laid down regulatory requirements. Yet the regulatory regimes in different States have evolved differently and therefore, has certain attributes which are unique to these and confer on them their identity.

  9. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  10. Level set method for optimal shape design of MRAM core. Micromagnetic approach

    International Nuclear Information System (INIS)

    Melicher, Valdemar; Cimrak, Ivan; Keer, Roger van

    2008-01-01

    We aim at optimizing the shape of the magnetic core in MRAM memories. The evolution of the magnetization during the writing process is described by the Landau-Lifshitz equation (LLE). The actual shape of the core in one cell is characterized by the coefficient γ. Cost functional f=f(γ) expresses the quality of the writing process having in mind the competition between the full-select and the half-select element. We derive an explicit form of the derivative F=∂f/∂γ which allows for the use of gradient-type methods for the actual computation of the optimized shape (e.g., steepest descend method). The level set method (LSM) is employed for the representation of the piecewise constant coefficient γ

  11. Optimizing a three-element core design for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    West, C.D.

    1995-01-01

    Source of neutrons in the proposed Advanced Neutron Source facility is a multipurpose research reactor providing 5-10 times the flux, for neutron beams, of the best existing facilities. Baseline design for the reactor core, based on the ''no new inventions'' rule, was an assembly of two annular fuel elements similar to those used in the Oak Ridge and Grenoble high flux reactors, containing highly enriched U silicide particles. DOE commissioned a study of the use of medium- or low-enriched U; a three-element core design was studied as a means to provide extra volume to accommodate the additional U compound required when the fissionable 235 U has to be diluted with 238 U to reduce the enrichment. This paper describes the design and optimization of that three-element core

  12. Nuclear Power Plant Outage Optimization Strategy. 2016 Edition

    International Nuclear Information System (INIS)

    2016-10-01

    This publication is an update of IAEA-TECDOC-1315, Nuclear Power Plant Outage Optimisation Strategy, which was published in 2002, and aims to communicate good outage management practices in a manner that can be used by operators and utilities in Member States. Nuclear power plant outage management is a key factor for safe and economic nuclear power plant performance. This publication discusses plant outage strategy and how this strategy is actually implemented. The main areas that are important for outage optimization that were identified by the utilities and government organizations participating in this report are: 1) organization and management; 2) outage planning and preparation; 3) outage execution; 4) safety outage review; and 5) counter measures to avoid the extension of outages and to facilitate the work in forced outages. Good outage management practices cover many different areas of work and this publication aims to communicate these good practices in a way that they can be used effectively by operators and utilities

  13. Dynamic Simulation and Optimization of Nuclear Hydrogen Production Systems

    Energy Technology Data Exchange (ETDEWEB)

    Paul I. Barton; Mujid S. Kaximi; Georgios Bollas; Patricio Ramirez Munoz

    2009-07-31

    This project is part of a research effort to design a hydrogen plant and its interface with a nuclear reactor. This project developed a dynamic modeling, simulation and optimization environment for nuclear hydrogen production systems. A hybrid discrete/continuous model captures both the continuous dynamics of the nuclear plant, the hydrogen plant, and their interface, along with discrete events such as major upsets. This hybrid model makes us of accurate thermodynamic sub-models for the description of phase and reaction equilibria in the thermochemical reactor. Use of the detailed thermodynamic models will allow researchers to examine the process in detail and have confidence in the accurary of the property package they use.

  14. The efficiency and fidelity of the in-core nuclear fuel management code FORMOSA-P

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1994-01-01

    The second-order generalized perturbation theory (GPT), nodal neutronic model utilized within the nuclear fuel management optimization code FORMOSA-P is presented within the context of prediction fidelity and computational efficiency versus forward solution. Key features of thr GPT neutronics model as implemented within the Simulated Annealing optimization adaptive control algorithm are discussed. Supporting results are then presented demonstrating the superior consistency of adaptive control for both global and local optimization searches. (authors). 15 refs., 1 fig., 4 tabs

  15. Preliminary study for a nuclear multi-cycle reload optimization system

    International Nuclear Information System (INIS)

    Baptista, Rafael Pereira; Lima, Alan Miranda M. de; Medeiros, Jose Antonio Carlos Canedo; Schirru, Roberto

    2007-01-01

    Fuel assemblies in a reactor are discharged normally after several fuel cycles. This happens because of the concentration of fissile material existing in the fuel assemblies in the core decreases to values such that it is not more possible to keep the reactor operating producing energy at normal rated power. Therefore, the refueling optimization for a nuclear power plant is in fact a multi-cycle problem. A typical multi-cycle reload optimization depends on several kinds of relationships: one is the relationship between the locations where the fuel assemblies are placed for a specified fuel cycle; another is the relationship between fuel loading patterns for the subsequent fuel cycles. This makes the problem very complex and difficult to solve. Until the moment, all the presented proposals for solution are far from solving the multi-cycle optimization problems in reactor fuel management. In this work, we will show preliminary studies of possible solutions for a typical multi-cycle reload optimization problem trying to consider most important restrictions of a real model. In the initial comparisons, the optimization results will be compared with those obtained by the successive single cycle optimizations. (author)

  16. Optimizing the Betts-Miller-Janjic cumulus parameterization with Intel Many Integrated Core (MIC) architecture

    Science.gov (United States)

    Huang, Melin; Huang, Bormin; Huang, Allen H.-L.

    2015-10-01

    The schemes of cumulus parameterization are responsible for the sub-grid-scale effects of convective and/or shallow clouds, and intended to represent vertical fluxes due to unresolved updrafts and downdrafts and compensating motion outside the clouds. Some schemes additionally provide cloud and precipitation field tendencies in the convective column, and momentum tendencies due to convective transport of momentum. The schemes all provide the convective component of surface rainfall. Betts-Miller-Janjic (BMJ) is one scheme to fulfill such purposes in the weather research and forecast (WRF) model. National Centers for Environmental Prediction (NCEP) has tried to optimize the BMJ scheme for operational application. As there are no interactions among horizontal grid points, this scheme is very suitable for parallel computation. With the advantage of Intel Xeon Phi Many Integrated Core (MIC) architecture, efficient parallelization and vectorization essentials, it allows us to optimize the BMJ scheme. If compared to the original code respectively running on one CPU socket (eight cores) and on one CPU core with Intel Xeon E5-2670, the MIC-based optimization of this scheme running on Xeon Phi coprocessor 7120P improves the performance by 2.4x and 17.0x, respectively.

  17. An intelligent stochastic optimization routine for nuclear fuel cycle design

    International Nuclear Information System (INIS)

    Parks, G.T.

    1990-01-01

    A simulated annealing (Metropolis algorithm) optimization routine named AMETROP, which has been developed for use on realistic nuclear fuel cycle problems, is introduced. Each stage of the algorithm is described and the means by which it overcomes or avoids the difficulties posed to conventional optimization routines by such problems are explained. Special attention is given to innovations that enhance AMETROP's performance both through artificial intelligence features, in which the routine uses the accumulation of data to influence its future actions, and through a family of simple performance aids, which allow the designer to use his heuristic knowledge to guide the routine's essentially random search. Using examples from a typical fuel cycle optimization problem, the performance of the stochastic Metropolis algorithm is compared to that of the only suitable deterministic routine in a standard software library, showing AMETROP to have many advantages

  18. Plant Design Nuclear Fuel Element Production Capacity Optimization to Support Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    Bambang Galung Susanto

    2007-01-01

    The optimization production capacity for designing nuclear fuel element fabrication plant in Indonesia to support the nuclear power plant has been done. From calculation and by assuming that nuclear power plant to be built in Indonesia as much as 12 NPP and having capacity each 1000 MW, the optimum capacity for nuclear fuel element fabrication plant is 710 ton UO 2 /year. The optimum capacity production selected, has considered some aspects such as fraction batch (cycle, n = 3), length of cycle (18 months), discharge burn-up value (Bd) 35,000 up 50,000 MWD/ton U, enriched uranium to be used in the NPP (3.22 % to 4.51 %), future market development for fuel element, and the trend of capacity production selected by advances country to built nuclear fuel element fabrication plant type of PWR. (author)

  19. Modelling of core protection and monitoring system for PWR nuclear power plant simulator

    International Nuclear Information System (INIS)

    Jung Kun Lee; Byoung Sung Han

    1997-01-01

    A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

  20. Study on in-core fuel management for CNP1500 nuclear power plant

    International Nuclear Information System (INIS)

    Li Dongsheng

    2005-10-01

    CNP1500 is a four-loop PWR nuclear power plant with light water as moderator and coolant. The reactor core is composed of 205 AFA-3GXL fuel assemblies. The active core height at cold is 426.4 cm and equivalent diameter is 347.0 cm. The reactor thermal output is 4250 MW, and average linear power density is 179.5 W/cm. The cycle length of equilibrium cycle core is 470 equivalent full power days. For all cycles, the moderator temperature coefficients at all conditions are negative values, the nuclear enthalpy rise factors F ΔH at hot full power, all control rods out and equilibrium xenon are less than the limit value, the maximum discharge assembly burnup is less 55000 MW·d/tU, and the shutdown margin values at the end of life meet design criteria. The low-leakage core loading reduces radiation damage on pressure vessel and is beneficial to prolong use lifetime of it. The in-core fuel management design scheme and main calculation results for CNP1500 nuclear power plant are presented. (author)

  1. NUCORE - A system for nuclear structure calculations with cluster-core models

    International Nuclear Information System (INIS)

    Heras, C.A.; Abecasis, S.M.

    1982-01-01

    Calculation of nuclear energy levels and their electromagnetic properties, modelling the nucleus as a cluster of a few particles and/or holes interacting with a core which in turn is modelled as a quadrupole vibrator (cluster-phonon model). The members of the cluster interact via quadrupole-quadrupole and pairing forces. (orig.)

  2. A new nodal kinetics method for analyzing fast control rod motions in nuclear reactor cores

    International Nuclear Information System (INIS)

    Kaya, S.; Yavuz, H.

    2001-01-01

    A new nodal kinetics approach is developed for analyzing large reactivity accidents in nuclear reactor cores. This method shows promising that it has capability of inspecting promt criticality transients and it gives comparable results with respect to those of other techniques. (orig.)

  3. Results of an analysis of in-core measurements during the first core cycle of the Greifswald nuclear power plant, unit 3

    International Nuclear Information System (INIS)

    Gehre, G.

    1982-01-01

    First results of an analysis of flux and temperature values obtained from the in-core system in the third unit of the Greifswald nuclear power plant during the first core cycle are presented. The analysis has been performed with the aid of the computer code INCA. Possibilities and limits of this code are shown. (author)

  4. Optimization of the radioprotection for nuclear medicine services

    International Nuclear Information System (INIS)

    Lira, Renata F. de; Filho, Joao A.; Santos, Luiz A.P.; Lima, Fernando Roberto de Andrade; Vieira, Jose W.

    2013-01-01

    Nuclear medicine (NM) is a medical specialty which uses small amounts of radioactive material combined with drugs, to make either therapeutic treatments or form diagnostic images of the organ and tissue. Follow the nuclear regulations any activity involving ionizing radiation should be justified and it must have their procedures of work to be optimized. Thus, the aim of the study is to determine the need and the importance of optimization of radiation protection in NM services and reduce occupationally exposed individuals (OEI) doses in order to avoid possible contamination or accidents and reduce the costs of protection. Optimization for a NM service that makes use of ionizing radiation can be performed using different techniques such as the expanded cost-benefit analysis. Such technique introduces one or two attributes associated to the detriment cost, Y, and the protection costs, X. This work was conducted in the year 2011, where it was analyzed data of 56 employees from 2002 to 2010. The value of the cost of protection was R$ 147.645,95, including accessories, courses, training and maintenance costs. On the other hand, the cost of the expense ranged from R$ 1.065.750, 00 up to R$ 28.890.351, 00 and the parameter responsible for this variation is the collective dose. The increasing of these dose values causes the increasing of the total costs, and one can conclude that there really is an importance of applying the optimization technique to improve the safety of OEI at the nuclear medicine service and reducing costs of protection. (author)

  5. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    Science.gov (United States)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  6. Optimization of refueling-shuffling scheme in PWR core by random search strategy

    International Nuclear Information System (INIS)

    Wu Yuan

    1991-11-01

    A random method for simulating optimization of refueling management in a pressurized water reactor (PWR) core is described. The main purpose of the optimization was to select the 'best' refueling arrangement scheme which would produce maximum economic benefits under certain imposed conditions. To fulfill this goal, an effective optimization strategy, two-stage random search method was developed. First, the search was made in a manner similar to the stratified sampling technique. A local optimum can be reached by comparison of the successive results. Then the other random experiences would be carried on between different strata to try to find the global optimum. In general, it can be used as a practical tool for conventional fuel management scheme. However, it can also be used in studies on optimization of Low-Leakage fuel management. Some calculations were done for a typical PWR core on a CYBER-180/830 computer. The results show that the method proposed can obtain satisfactory approach at reasonable low computational cost

  7. Uncertainly propagation analysis for Yonggwang nuclear unit 4 by McCARD/MASTER core analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Dong Hyuk; Shim, Hyung Jin; Kim, Chang Hyo [Seoul National University, Seoul (Korea, Republic of)

    2014-06-15

    This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor (k{sub eff}), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

  8. Development of Reactor Core Model based on Optimal Analysis for Shinhanul no. 1, 2 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-min [Korea Hydro Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    As one of the outputs of 'Development of the Shin Hanul Nuclear Plant(SHN) 1,2 Simulator' project which is being done by KHNP Central Research Institute, the SHN1,2 Simulator is being developed including the KNICS methodology and advanced Alarm Systems first applied to the Nuclear Power Plant in Korea, and the SHN 1,2 simulator adopts the virtually stimulated HMI(Human-Machine Interface) for the non-safety MMIS system, whose key-programs are identical to those applied to the real SHN 1,2 plants. The purpose of this paper is to develop localization core model by integrating the Simulator system with the Simulator core model though technology agreement of KAERI. To develop ShinHanul 1 and 2 reactor core simulator model, KHNP and KAERI create MASTER-SIM model and tried validation. And calculations of MASSIM{sub S}S program for MASTER{sub S}IM validation, are within tolerance range. Test has not yet been completed. And many verification will be conducted MASTER-SIM software is expected to be the highest economic software and satisfy international simulator standards.

  9. Rules for design of nuclear graphite core components - some considerations and approaches

    International Nuclear Information System (INIS)

    Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

    1978-01-01

    The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)

  10. Phase diagram of nuclear 'pasta' and its uncertainties in supernova cores

    International Nuclear Information System (INIS)

    Sonoda, Hidetaka; Watanabe, Gentaro; Sato, Katsuhiko; Yasuoka, Kenji; Ebisuzaki, Toshikazu

    2008-01-01

    We examine the model dependence of the phase diagram of inhomogeneous nulcear matter in supernova cores using the quantum molecular dynamics (QMD). Inhomogeneous matter includes crystallized matter with nonspherical nuclei--''pasta'' phases--and the liquid-gas phase-separating nuclear matter. Major differences between the phase diagrams of the QMD models can be explained by the energy of pure neutron matter at low densities and the saturation density of asymmetric nuclear matter. We show the density dependence of the symmetry energy is also useful to understand uncertainties of the phase diagram. We point out that, for typical nuclear models, the mass fraction of the pasta phases in the later stage of the collapsing cores is higher than 10-20%

  11. Protein kinases responsible for the phosphorylation of the nuclear egress core complex of human cytomegalovirus.

    Science.gov (United States)

    Sonntag, Eric; Milbradt, Jens; Svrlanska, Adriana; Strojan, Hanife; Häge, Sigrun; Kraut, Alexandra; Hesse, Anne-Marie; Amin, Bushra; Sonnewald, Uwe; Couté, Yohann; Marschall, Manfred

    2017-10-01

    Nuclear egress of herpesvirus capsids is mediated by a multi-component nuclear egress complex (NEC) assembled by a heterodimer of two essential viral core egress proteins. In the case of human cytomegalovirus (HCMV), this core NEC is defined by the interaction between the membrane-anchored pUL50 and its nuclear cofactor, pUL53. NEC protein phosphorylation is considered to be an important regulatory step, so this study focused on the respective role of viral and cellular protein kinases. Multiply phosphorylated pUL50 varieties were detected by Western blot and Phos-tag analyses as resulting from both viral and cellular kinase activities. In vitro kinase analyses demonstrated that pUL50 is a substrate of both PKCα and CDK1, while pUL53 can also be moderately phosphorylated by CDK1. The use of kinase inhibitors further illustrated the importance of distinct kinases for core NEC phosphorylation. Importantly, mass spectrometry-based proteomic analyses identified five major and nine minor sites of pUL50 phosphorylation. The functional relevance of core NEC phosphorylation was confirmed by various experimental settings, including kinase knock-down/knock-out and confocal imaging, in which it was found that (i) HCMV core NEC proteins are not phosphorylated solely by viral pUL97, but also by cellular kinases; (ii) both PKC and CDK1 phosphorylation are detectable for pUL50; (iii) no impact of PKC phosphorylation on NEC functionality has been identified so far; (iv) nonetheless, CDK1-specific phosphorylation appears to be required for functional core NEC interaction. In summary, our findings provide the first evidence that the HCMV core NEC is phosphorylated by cellular kinases, and that the complex pattern of NEC phosphorylation has functional relevance.

  12. Kriging-based algorithm for nuclear reactor neutronic design optimization

    International Nuclear Information System (INIS)

    Kempf, Stephanie; Forget, Benoit; Hu, Lin-Wen

    2012-01-01

    Highlights: ► A Kriging-based algorithm was selected to guide research reactor optimization. ► We examined impacts of parameter values upon the algorithm. ► The best parameter values were incorporated into a set of best practices. ► Algorithm with best practices used to optimize thermal flux of concept. ► Final design produces thermal flux 30% higher than other 5 MW reactors. - Abstract: Kriging, a geospatial interpolation technique, has been used in the present work to drive a search-and-optimization algorithm which produces the optimum geometric parameters for a 5 MW research reactor design. The technique has been demonstrated to produce an optimal neutronic solution after a relatively small number of core calculations. It has additionally been successful in producing a design which significantly improves thermal neutron fluxes by 30% over existing reactors of the same power rating. Best practices for use of this algorithm in reactor design were identified and indicated the importance of selecting proper correlation functions.

  13. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  14. Optimal Procedure for siting of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Aziuddin, Khairiah Binti; Park, Seo Yeon; Roh, Myung Sub

    2013-01-01

    This study discusses on a simulation approach for sensitivity analysis of the weights of multi-criteria decision models. The simulation procedures can also be used to aid the actual decision process, particularly when the task is to select a subset of superior alternatives. This study is to identify the criteria or parameters which are sensitive to the weighting factor that can affect the results in the decision making process to determine the optimal site for nuclear power plant (NPP) site. To perform this study, we adhere to IAEA NS-R-3 and DS 433. The siting process for nuclear installation consists of site survey and site selection stages. The siting process generally consists of an investigation of a large region to select one or more candidate sites by surveying the sites. After comparing the ROI, two candidate sites are compared for final determination, which are Wolsong and Kori site. Some assumptions are taken into consideration due to limitations and constraints throughout performing this study. Sensitivity analysis of multi criteria decision models is performed in this study to determine the optimal site in the site selection stage. Logical Decisions software will be employed as a tool to perform this analysis. Logical Decisions software helps to formulate the preferences and then rank the alternatives. It provides clarification of the rankings and hence aids the decision makers on evaluating the alternatives, and finally draw a conclusion on the selection of the optimal site

  15. Optimal Procedure for siting of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Aziuddin, Khairiah Binti; Park, Seo Yeon; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    This study discusses on a simulation approach for sensitivity analysis of the weights of multi-criteria decision models. The simulation procedures can also be used to aid the actual decision process, particularly when the task is to select a subset of superior alternatives. This study is to identify the criteria or parameters which are sensitive to the weighting factor that can affect the results in the decision making process to determine the optimal site for nuclear power plant (NPP) site. To perform this study, we adhere to IAEA NS-R-3 and DS 433. The siting process for nuclear installation consists of site survey and site selection stages. The siting process generally consists of an investigation of a large region to select one or more candidate sites by surveying the sites. After comparing the ROI, two candidate sites are compared for final determination, which are Wolsong and Kori site. Some assumptions are taken into consideration due to limitations and constraints throughout performing this study. Sensitivity analysis of multi criteria decision models is performed in this study to determine the optimal site in the site selection stage. Logical Decisions software will be employed as a tool to perform this analysis. Logical Decisions software helps to formulate the preferences and then rank the alternatives. It provides clarification of the rankings and hence aids the decision makers on evaluating the alternatives, and finally draw a conclusion on the selection of the optimal site.

  16. Optimized application of systems engineering to nuclear waste repository projects

    International Nuclear Information System (INIS)

    Miskimin, P.A.; Shepard, M.

    1986-01-01

    The purpose of this presentation is to describe a fully optimized application of systems engineering methods and philosophy to the management of a large nuclear waste repository project. Knowledge gained from actual experience with the use of the systems approach on two repository projects is incorporated in the material presented. The projects are currently evaluating the isolation performance of different geologic settings and are in different phases of maturity. Systems engineering methods were applied by the principal author at the Waste Isolation Pilot Plant (WIPP) in the form of a functional analysis. At the Basalt Waste Isolation Project (BWIP), the authors assisted the intergrating contractor with the development and application of systems engineering methods. Based on this experience and that acquired from other waste management projects, an optimized plan for applying systems engineering techniques was developed. The plan encompasses the following aspects: project organization, developing and defining requirements, assigning work responsibilities, evaluating system performance, quality assurance, controlling changes, enhancing licensability, optimizing project performance, and addressing regulatory issues. This information is presented in the form of a roadmap for the practical application of system engineering principles to a nuclear waste repository project

  17. IP cores design from specifications to production modeling, verification, optimization, and protection

    CERN Document Server

    Mohamed, Khaled Salah

    2016-01-01

    This book describes the life cycle process of IP cores, from specification to production, including IP modeling, verification, optimization, and protection. Various trade-offs in the design process are discussed, including  those associated with many of the most common memory cores, controller IPs  and system-on-chip (SoC) buses. Readers will also benefit from the author’s practical coverage of new verification methodologies. such as bug localization, UVM, and scan-chain.  A SoC case study is presented to compare traditional verification with the new verification methodologies. ·         Discusses the entire life cycle process of IP cores, from specification to production, including IP modeling, verification, optimization, and protection; ·         Introduce a deep introduction for Verilog for both implementation and verification point of view.  ·         Demonstrates how to use IP in applications such as memory controllers and SoC buses. ·         Describes a new ver...

  18. Behavior-aware cache hierarchy optimization for low-power multi-core embedded systems

    Science.gov (United States)

    Zhao, Huatao; Luo, Xiao; Zhu, Chen; Watanabe, Takahiro; Zhu, Tianbo

    2017-07-01

    In modern embedded systems, the increasing number of cores requires efficient cache hierarchies to ensure data throughput, but such cache hierarchies are restricted by their tumid size and interference accesses which leads to both performance degradation and wasted energy. In this paper, we firstly propose a behavior-aware cache hierarchy (BACH) which can optimally allocate the multi-level cache resources to many cores and highly improved the efficiency of cache hierarchy, resulting in low energy consumption. The BACH takes full advantage of the explored application behaviors and runtime cache resource demands as the cache allocation bases, so that we can optimally configure the cache hierarchy to meet the runtime demand. The BACH was implemented on the GEM5 simulator. The experimental results show that energy consumption of a three-level cache hierarchy can be saved from 5.29% up to 27.94% compared with other key approaches while the performance of the multi-core system even has a slight improvement counting in hardware overhead.

  19. Plant life management optimized utilization of existing nuclear power plants

    International Nuclear Information System (INIS)

    Watzinger, H.; Erve, M.

    1999-01-01

    For safe, reliable and economical nuclear power generation it is of central importance to understand, analyze and manage aging-related phenomena and to apply this information in the systematic utilization and as-necessary extension of the service life of components and systems. An operator's overall approach to aging and plant life management which also improves performance characteristics can help to optimize plant operating economy. In view of the deregulation of the power generation industry with its increased competition, nuclear power plants must today also increasingly provide for or maintain a high level of plant availability and low power generating costs. This is a difficult challenge even for the newest, most modern plants, and as plants age they can only remain competitive if a plant operator adopts a strategic approach which takes into account the various aging-related effects on a plant-wide basis. The significance of aging and plant life management for nuclear power plants becomes apparent when looking at their age: By the year 2000 roughly fifty of the world's 434 commercial nuclear power plants will have been in operation for thirty years or more. According to the International Atomic Energy Agency, as many as 110 plants will have reached the thirty-year service mark by the year 2005. In many countries human society does not push the construction of new nuclear power plants and presumably will not change mind within the next ten years. New construction licenses cannot be expected so that for economical and ecological reasons existing plants have to be operated unchallengeably. On the other hand the deregulation of the power production market is asking just now for analysis of plant life time to operate the plants at a high technical and economical level until new nuclear power plants can be licensed and constructed. (author)

  20. The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

    International Nuclear Information System (INIS)

    Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

    2009-01-01

    The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

  1. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  2. Risk Informed Optimization of Nuclear Instrumentation for Emergency Conditions

    International Nuclear Information System (INIS)

    Knoll, Alexander

    2013-06-01

    Emergency conditions after a nuclear accident are different in each case and cannot be predicted accurately. The accident at TMI did not contaminate the environment. The accident at Chernobyl had a large, early release of nuclear contamination, widely dispersed over many countries. Although there was no large, early release of contamination at Fukushima Daiichi, the timeline of the accident included days of later contamination of various degrees of severity. A large amount of the contamination has been released to the ocean and an exclusion zone still exists around the station. In all of these accidents there were no adequate radiation monitoring systems distant from the origin point that could provide accurate status to the authorities and the local population. In the recent years a number of new nuclear monitoring systems have been implemented or are under development to be installed in areas that might be exposed to nuclear contamination in emergencies. Based on the risk informed optimization methodology, this paper provides recommendations for selecting the quantity and type of instrumentation, the location and sampling of data, and the real-time processing of information. (authors)

  3. Evaluating the fundamental qualities of a nuclear medicine radiographer for the provision of an optimal clinical service

    International Nuclear Information System (INIS)

    Griffiths, Marc; King, Simon; Stewart, Rob; Dawson, Gary

    2010-01-01

    The developing nature of nuclear medicine practice highlights the need for an evaluation of the fundamental qualities of a Radiographer working within this discipline. Existing guidelines appear to be in place for clinical technologists working within nuclear medicine. However, limited guidance has been provided for Radiographers practicing within this discipline. This article aims to discuss the fundamental qualities that are considered essential for optimal service delivery, following consultation with various stakeholders. Areas such as technical expertise and knowledge, appropriate use of imaging equipment and current models of safe working practice will be discussed. Patient care and ethical considerations will also be evaluated, along with some core recommendations for future advanced practice.

  4. Nuclear fuel cycle optimization - methods and modelling techniques

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1982-01-01

    This book is aimed at presenting methods applicable in the analysis of fuel cycle logistics and optimization as well as in evaluating the economics of different reactor strategies. After a succinct introduction to the phases of a fuel cycle, uranium cost trends are assessed in a global perspective and subsequent chapters deal with the fuel cycle problems faced by a power utility. A fundamental material flow model is introduced first in the context of light water reactor fuel cycles. Besides the minimum cost criterion, the text also deals with other objectives providing for a treatment of cost uncertainties and of the risk of proliferation of nuclear weapons. Methods to assess mixed reactor strategies, comprising also other reactor types than the light water reactor, are confined to cost minimization. In the final Chapter, the integration of nuclear capacity within a generating system is examined. (author)

  5. Optimal design of condenser volume in nuclear power plant

    International Nuclear Information System (INIS)

    Zheng Jing; Yan Changqi; Wang Jianjun

    2011-01-01

    The condenser is an important component in the nuclear power plant,whose dimension will influence the economy and the arrangement of the nuclear power plant.In this paper, the calculation model was established according to the design experience. The corresponding codes were also developed. The sensitivity of design parameters which influence the condenser Janume was analyzed. The present optimal design of the condenser, aiming at the volume minimization, was carried out with the self-developed complex-genetic algorithm. The results show that the reference condenser design is far from the best scheme. In addition, the results also verify the feasibility of the complex-genetic algorithm. Furthermore, the results of this paper can provide reference for the design of the condenser. (authors)

  6. A discrete optimization method for nuclear fuel management

    International Nuclear Information System (INIS)

    Argaud, J.P.

    1993-04-01

    Nuclear loading pattern elaboration can be seen as a combinational optimization problem, of tremendous size and with non-linear cost-functions, and search are always numerically expensive. After a brief introduction of the main aspects of nuclear fuel management, this note presents a new idea to treat the combinational problem by using informations included in the gradient of a cost function. The method is to choose, by direct observation of the gradient, the more interesting changes in fuel loading patterns. An example is then developed to illustrate an operating mode of the method, and finally, connections with simulated annealing and genetic algorithms are described as an attempt to improve search processes. (author). 1 fig., 16 refs

  7. Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

    1990-01-01

    It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

  8. Dose reduction and optimization studies (ALARA) at nuclear power facilities

    International Nuclear Information System (INIS)

    Baum, J.W.; Meinhold, C.B.

    1983-01-01

    Brookhaven National Laboratory (BNL) has been commissioned by the Nuclear Regulatory Commission (NRC) to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at LWR plants. These studies have the following objectives: identify high-dose maintenance tasks; identify dose-reduction techniques; examine incentives for dose reduction; evaluate cost-effectiveness and optimization of dose-reduction techniques; and compile an ALARA handbook on data, engineering modifications, cost-effectiveness calculations, and other information of interest to ALARA practioners

  9. Optimizing Performance of Combustion Chemistry Solvers on Intel's Many Integrated Core (MIC) Architectures

    Energy Technology Data Exchange (ETDEWEB)

    Sitaraman, Hariswaran [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Grout, Ray W [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-06-09

    This work investigates novel algorithm designs and optimization techniques for restructuring chemistry integrators in zero and multidimensional combustion solvers, which can then be effectively used on the emerging generation of Intel's Many Integrated Core/Xeon Phi processors. These processors offer increased computing performance via large number of lightweight cores at relatively lower clock speeds compared to traditional processors (e.g. Intel Sandybridge/Ivybridge) used in current supercomputers. This style of processor can be productively used for chemistry integrators that form a costly part of computational combustion codes, in spite of their relatively lower clock speeds. Performance commensurate with traditional processors is achieved here through the combination of careful memory layout, exposing multiple levels of fine grain parallelism and through extensive use of vendor supported libraries (Cilk Plus and Math Kernel Libraries). Important optimization techniques for efficient memory usage and vectorization have been identified and quantified. These optimizations resulted in a factor of ~ 3 speed-up using Intel 2013 compiler and ~ 1.5 using Intel 2017 compiler for large chemical mechanisms compared to the unoptimized version on the Intel Xeon Phi. The strategies, especially with respect to memory usage and vectorization, should also be beneficial for general purpose computational fluid dynamics codes.

  10. A Bandwidth-Optimized Multi-Core Architecture for Irregular Applications

    Energy Technology Data Exchange (ETDEWEB)

    Secchi, Simone; Tumeo, Antonino; Villa, Oreste

    2012-05-31

    This paper presents an architecture template for next-generation high performance computing systems specifically targeted to irregular applications. We start our work by considering that future generation interconnection and memory bandwidth full-system numbers are expected to grow by a factor of 10. In order to keep up with such a communication capacity, while still resorting to fine-grained multithreading as the main way to tolerate unpredictable memory access latencies of irregular applications, we show how overall performance scaling can benefit from the multi-core paradigm. At the same time, we also show how such an architecture template must be coupled with specific techniques in order to optimize bandwidth utilization and achieve the maximum scalability. We propose a technique based on memory references aggregation, together with the related hardware implementation, as one of such optimization techniques. We explore the proposed architecture template by focusing on the Cray XMT architecture and, using a dedicated simulation infrastructure, validate the performance of our template with two typical irregular applications. Our experimental results prove the benefits provided by both the multi-core approach and the bandwidth optimization reference aggregation technique.

  11. Nuclear-fuel-cycle optimization: methods and modelling techniques

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1982-01-01

    This book present methods applicable to analyzing fuel-cycle logistics and optimization as well as in evaluating the economics of different reactor strategies. After an introduction to the phases of a fuel cycle, uranium cost trends are assessed in a global perspective. Subsequent chapters deal with the fuel-cycle problems faced by a power utility. The fuel-cycle models cover the entire cycle from the supply of uranium to the disposition of spent fuel. The chapter headings are: Nuclear Fuel Cycle, Uranium Supply and Demand, Basic Model of the LWR (light water reactor) Fuel Cycle, Resolution of Uncertainties, Assessment of Proliferation Risks, Multigoal Optimization, Generalized Fuel-Cycle Models, Reactor Strategy Calculations, and Interface with Energy Strategies. 47 references, 34 figures, 25 tables

  12. Exergy analysis for Generation IV nuclear plant optimization

    International Nuclear Information System (INIS)

    Gomez, A.; Azzaro-Pantel, C.; Domenech, S.; Pibouleau, L.; Latge, Ch.; Haubensack, D.; Dumaz, P.

    2010-01-01

    This paper deals with the application of the exergy concept to an energy production system involving a very high temperature reactor coupled with an innovative electricity-generating cycle. The objective is to propose a general approach to quantify exergy destruction of the involved process components, modelled by a thermodynamic simulator (Proceedings of the Conference on High Temperature Reactors, Beijing, China, 22-24 September 2004, International Atomic Agency, Vienna (Austria), HTR-2004; 1-11). The minimization of exergy destruction is then identified as the optimization criterion used in an optimization framework based on a genetic algorithm, in which the model is embedded. Finally, the approach is applied to electrical production by a Brayton-Rankine combined cycle connected to a nuclear reactor. Some typical results are presented. The perspectives of this work including the cogeneration of hydrogen and electricity are highlighted. (authors)

  13. Exergy analysis for Generation IV nuclear plant optimization

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, A.; Azzaro-Pantel, C.; Domenech, S.; Pibouleau, L. [Univ Toulouse, Lab Genie Chim, CNRS, UMR 5503, F-31700 Toulouse 1 (France); Latge, Ch. [CEA Cadarache DEN DTN DIR, St Paul Les Durance, (France); Haubensack, D.; Dumaz, P. [CEA Cadarache DEN DER SESI LCSI, St Paul Les Durance (France)

    2010-07-01

    This paper deals with the application of the exergy concept to an energy production system involving a very high temperature reactor coupled with an innovative electricity-generating cycle. The objective is to propose a general approach to quantify exergy destruction of the involved process components, modelled by a thermodynamic simulator (Proceedings of the Conference on High Temperature Reactors, Beijing, China, 22-24 September 2004, International Atomic Agency, Vienna (Austria), HTR-2004; 1-11). The minimization of exergy destruction is then identified as the optimization criterion used in an optimization framework based on a genetic algorithm, in which the model is embedded. Finally, the approach is applied to electrical production by a Brayton-Rankine combined cycle connected to a nuclear reactor. Some typical results are presented. The perspectives of this work including the cogeneration of hydrogen and electricity are highlighted. (authors)

  14. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  15. Core management and fuel handling for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    This Safety Guide supplements and elaborates upon the safety requirements for core management and fuel handling that are presented in Section 5 of the Safety Requirements publication on the operation of nuclear power plants. The present publication supersedes the IAEA Safety Guide on Safety Aspects of Core Management and Fuel Handling, issued in 1985 as Safety Series No. 50-SG-010. It is also related to the Safety Guide on the Operating Organization for Nuclear Power Plants, which identifies fuel management as one of the various functions to be performed by the operating organization. The purpose of this Safety Guide is to provide recommendations for core management and fuel handling at nuclear power plants on the basis of current international good practice. The present Safety Guide addresses those aspects of fuel management activities that are necessary in order to allow optimum reactor core operation without compromising the limits imposed by the design safety considerations relating to the nuclear fuel and the plant as a whole. In this publication, 'core management' refers to those activities that are associated with fuel management in the core and reactivity control, and 'fuel handling' refers to the movement, storage and control of fresh and irradiated fuel. Fuel management comprises both core management and fuel handling. This Safety Guide deals with fuel management for all types of land based stationary thermal neutron power plants. It describes the safety objectives of core management, the tasks that have to be accomplished to meet these objectives and the activities undertaken to perform those tasks. It also deals with the receipt of fresh fuel, storage and handling of fuel and other core components, the loading and unloading of fuel and core components, and the insertion and removal of other reactor materials. In addition, it deals with loading a transport container with irradiated fuel and its preparation for transport off the site. Transport

  16. Basic evaluation on nuclear characteristics of BWR high burnup MOX fuel and core

    International Nuclear Information System (INIS)

    Nagano, M.; Sakurai, S.; Yamaguchi, H.

    1997-01-01

    MOX fuel will be used in existing commercial BWR cores as a part of reload fuels with equivalent operability, safety and economy to UO 2 fuel in Japan. The design concept should be compatible with UO 2 fuel design. High burnup UO 2 fuels are being developed and commercialized step by step. The MOX fuel planned to be introduced in around year 2000 will use the same hardware as UO 2 8 x 8 array fuel developed for a second step of UO 2 high burnup fuel. The target discharge exposure of this MOX fuel is about 33 GWd/t. And the loading fraction of MOX fuel is approximately one-third in an equilibrium core. On the other hand, it becomes necessary to minimize a number of MOX fuels and plants utilizing MOX fuel, mainly due to the fuel economy, handling cost and inspection cost in site. For the above reasons, it needed to developed a high burnup MOX fuel containing much Pu and a core with a large amount of MOX fuels. The purpose of this study is to evaluate basic nuclear fuel and core characteristics of BWR high burnup MOX fuel with batch average exposure of about 39.5 GWd/t using 9 x 9 array fuel. The loading fraction of MOX fuel in the core is within a range of about 50% to 100%. Also the influence of Pu isotopic composition fluctuations and Pu-241 decay upon nuclear characteristics are studied. (author). 3 refs, 5 figs, 3 tabs

  17. Optimization design of toroidal core for magnetic energy harvesting near power line by considering saturation effect

    Science.gov (United States)

    Park, Bumjin; Kim, Dongwook; Park, Jaehyoung; Kim, Kibeom; Koo, Jay; Park, HyunHo; Ahn, Seungyoung

    2018-05-01

    Recently, magnetic energy harvesting technologies have been studied actively for self-sustainable operation of applications around power line. However, magnetic energy harvesting around power lines has the problem of magnetic saturation, which can cause power performance degradation of the harvester. In this paper, optimal design of a toroidal core for magnetic energy harvesters has been proposed with consideration of magnetic saturation near power lines. Using Permeability-H curve and Ampere's circuital law, the optimum dimensional parameters needed to generate induced voltage were analyzed via calculation and simulation. To reflect a real environment, we consider the nonlinear characteristic of the magnetic core material and supply current through a 3-phase distribution panel used in the industry. The effectiveness of the proposed design methodology is verified by experiments in a power distribution panel and takes 60.9 V from power line current of 60 A at 60 Hz.

  18. Demountable toroidal fusion core facility for physics optimization and fusion engineering

    International Nuclear Information System (INIS)

    Bogart, S.L.; Wagner, C.E.; Krall, N.A.; Dalessandro, J.A.; Weggel, C.F.; Lund, K.O.; Sedehi, S.

    1986-01-01

    Following a successful compact ignition tokamak (CIT) experiment, a fusion facility will be required for physics optimization (POF) and fusion engineering research (FERF). The POF will address issues such as high-beta operation, current drive, impurity control, and will test geometric and configurational variations such as the spherical torus or the reversed-field pinch (RFP). The FERF will be designed to accumulate rapidly a large neutron dose in prototypical fusion subsystems exposed to radiation. Both facilities will require low-cost replacement cores and rapid replacement times. The Demountable Toroidal Fusion Core (DTFC) facility is designed to fulfill these requirements. It would be a cost-effective stepping stone between the CIT and a demonstration fusion reactor

  19. Intelligent stochastic optimization routine for in-core fuel cycle design

    International Nuclear Information System (INIS)

    Parks, G.T.

    1988-01-01

    Any reactor fuel management strategy must specify the fuel design, batch sizes, loading configurations, and operational procedures for each cycle. To permit detailed design studies, the complex core characteristics must necessarily be computer modeled. Thus, the identification of an optimal fuel cycle design represents an optimization problem with a nonlinear objective function (OF), nonlinear safety constraints, many control variables, and no direct derivative information. Most available library routines cannot tackle such problems; this paper introduces an intelligent stochastic optimization routine that can. There has been considerable interest recently in the application of stochastic methods to difficult optimization problems, based on the statistical mechanics algorithms originally attributed to Metropolis. Previous work showed that, in optimizing the performance of a British advanced gas-cooled reactor fuel stringer, a rudimentary version of the Metropolis algorithm performed as efficiently as the only suitable routine in the Numerical Algorithms Group library. Since then the performance of the Metropolis algorithm has been considerably enhanced by the introduction of self-tuning capabilities by which the routine adjusts its control parameters and search pattern as it progresses. Both features can be viewed as examples of artificial intelligence, in which the routine uses the accumulation of data, or experience, to guide its future actions

  20. Multiobjective optimization for nuclear fleet evolution scenarios using COSI

    Directory of Open Access Journals (Sweden)

    Freynet David

    2016-01-01

    Full Text Available The consequences of various fleet evolution options on material inventories and flux in fuel cycle and waste can be analysed by means of transition scenario studies. The COSI code is currently simulating chronologically scenarios whose parameters are fully defined by the user and is coupled with the CESAR depletion code. As the interactions among reactors and fuel cycle facilities can be complex, and the ways in which they may be configured are many, the development of optimization methodology could improve scenario studies. The optimization problem definition needs to list: (i criteria (e.g. saving natural resources and minimizing waste production; (ii variables (scenario parameters related to reprocessing, reactor operation, installed power distribution, etc.; (iii constraints making scenarios industrially feasible. The large number of scenario calculations needed to solve an optimization problem can be time-consuming and hardly achievable; therefore, it requires the shortening of the COSI computation time. Given that CESAR depletion calculations represent about 95% of this computation time, CESAR surrogate models have been developed and coupled with COSI. Different regression models are compared to estimate CESAR outputs: first- and second-order polynomial regressions, Gaussian process and artificial neural network. This paper is about a first optimization study of a transition scenario from the current French nuclear fleet to a Sodium Fast Reactors fleet as defined in the frame of the 2006 French Act for waste management. The present article deals with obtaining the optimal scenarios and validating the methodology implemented, i.e. the coupling between the simulation software COSI, depletion surrogate models and a genetic algorithm optimization method.

  1. Optimal pin enrichment distributions in nuclear reactor fuel bundles

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1976-01-01

    A methodology has been developed to determine the fuel pin enrichment distribution that yields the best approximation to a prescribed power distribution in nuclear reactor fuel bundles. The problem is formulated as an optimization problem in which the optimal pin enrichments minimize the sum of squared deviations between the actual and prescribed fuel pin powers. A constant average enrichment constraint is imposed to ensure that a suitable value of reactivity is present in the bundle. When constraints are added that limit the fuel pins to a few enrichment types, one must determine not only the optimal values of the enrichment types but also the optimal distribution of the enrichment types amongst the pins. A matrix of boolean variables is used to describe the assignment of enrichment types to the pins. This nonlinear mixed integer programming problem may be rigorously solved with either exhaustive enumeration or branch and bound methods using a modification of the algorithm from the continuous problem as a suboptimization. Unfortunately these methods are extremely cumbersome and computationally overwhelming. Solutions which require only a moderate computational effort are obtained by assuming that the fuel pin enrichments in this problem are ordered as in the solution to the continuous problem. Under this assumption search schemes using either exhaustive enumeration or branch and bound become computationally attractive. An adaptation of the Hooke--Jeeves pattern search technique is shown to be especially efficient

  2. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

    International Nuclear Information System (INIS)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects

  3. Fundamental design bases for independent core cooling in Swedish nuclear power reactors

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

    New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

  4. Automated in-core image generation from video to aid visual inspection of nuclear power plant cores

    Energy Technology Data Exchange (ETDEWEB)

    Murray, Paul, E-mail: paul.murray@strath.ac.uk [Department of Electronic and Electrical Engineering, University of Strathclyde, Technology and Innovation Centre, 99 George Street, Glasgow, G1 1RD (United Kingdom); West, Graeme; Marshall, Stephen; McArthur, Stephen [Dept. Electronic and Electrical Engineering, University of Strathclyde, Royal College Building, 204 George Street, Glasgow G1 1XW (United Kingdom)

    2016-04-15

    Highlights: • A method is presented which improves visual inspection of reactor cores. • Significant time savings are made to activities on the critical outage path. • New information is extracted from existing data sources without additional overhead. • Examples from industrial case studies across the UK fleet of AGR stations. - Abstract: Inspection and monitoring of key components of nuclear power plant reactors is an essential activity for understanding the current health of the power plant and ensuring that they continue to remain safe to operate. As the power plants age, and the components degrade from their initial start-of-life conditions, the requirement for more and more detailed inspection and monitoring information increases. Deployment of new monitoring and inspection equipment on existing operational plant is complex and expensive, as the effect of introducing new sensing and imaging equipment to the existing operational functions needs to be fully understood. Where existing sources of data can be leveraged, the need for new equipment development and installation can be offset by the development of advanced data processing techniques. This paper introduces a novel technique for creating full 360° panoramic images of the inside surface of fuel channels from in-core inspection footage. Through the development of this technique, a number of technical challenges associated with the constraints of using existing equipment have been addressed. These include: the inability to calibrate the camera specifically for image stitching; dealing with additional data not relevant to the panorama construction; dealing with noisy images; and generalising the approach to work with two different capture devices deployed at seven different Advanced Gas Cooled Reactor nuclear power plants. The resulting data processing system is currently under formal assessment with a view to replacing the existing manual assembly of in-core defect montages. Deployment of the

  5. On the optimization of thorium bundle distribution in the initial core loading for a PHWR

    International Nuclear Information System (INIS)

    Mishra, S.; Ray, S.; Kumar, A.N.; Modak, R.S.; Ganesan, S.

    2009-01-01

    If the initial core of Indian 220 MWe PHWR is loaded with all fresh Natural Uranium fuel, only about 70% Full Power can be drawn in the initial operation due to large power peaking. It is possible to load few tens of Th bundles at selected locations to get nearly full power without violating any safety limits. Finding the best possible locations is, however, a fairly complex and massive combinatorial optimization problem. Here, optimum solutions are obtained by a latest evolutionary algorithm called EDA implemented on the EKA built at Computational Research Laboratories (CRL) in Pune. The effect of varying the number of Th bundles on results is discussed. (author)

  6. Parameters optimization defined by statistical analysis for cysteine-dextran radiolabeling with technetium tricarbonyl core.

    Science.gov (United States)

    Núñez, Eutimio Gustavo Fernández; Faintuch, Bluma Linkowski; Teodoro, Rodrigo; Wiecek, Danielle Pereira; da Silva, Natanael Gomes; Papadopoulos, Minas; Pelecanou, Maria; Pirmettis, Ioannis; de Oliveira Filho, Renato Santos; Duatti, Adriano; Pasqualini, Roberto

    2011-04-01

    The objective of this study was the development of a statistical approach for radiolabeling optimization of cysteine-dextran conjugates with Tc-99m tricarbonyl core. This strategy has been applied to the labeling of 2-propylene-S-cysteine-dextran in the attempt to prepare a new class of tracers for sentinel lymph node detection, and can be extended to other radiopharmaceuticals for different targets. The statistical routine was based on three-level factorial design. Best labeling conditions were achieved. The specific activity reached was 5 MBq/μg. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

  7. Parameters optimization defined by statistical analysis for cysteine-dextran radiolabeling with technetium tricarbonyl core

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez Nunez, Eutimio Gustavo, E-mail: eutimiocu@yahoo.co [Radiopharmacy Center, Institute of Energetic and Nuclear Research, Sao Paulo, SP 05508-000 (Brazil); Linkowski Faintuch, Bluma; Teodoro, Rodrigo; Pereira Wiecek, Danielle; Gomes da Silva, Natanael [Radiopharmacy Center, Institute of Energetic and Nuclear Research, Sao Paulo, SP 05508-000 (Brazil); Papadopoulos, Minas [Institute of Radioisotopes, Radiodiagnostic Products, National Center for Scientific Research ' Demokritos' , Athens (Greece); Pelecanou, Maria [Institute of Biology, National Center for Scientific Research ' Demokritos' , Athens (Greece); Pirmettis, Ioannis [Institute of Radioisotopes, Radiodiagnostic Products, National Center for Scientific Research ' Demokritos' , Athens (Greece); Santos Oliveira Filho, Renato de [Faculty of Medicine, Federal University of Sao Paulo, SP (Brazil); Duatti, Adriano [Department of Radiological Sciences, University of Ferrara, Ferrara (Italy); Pasqualini, Roberto [CIS Bio International, Gif sur Yvette (France)

    2011-04-15

    The objective of this study was the development of a statistical approach for radiolabeling optimization of cysteine-dextran conjugates with Tc-99m tricarbonyl core. This strategy has been applied to the labeling of 2-propylene-S-cysteine-dextran in the attempt to prepare a new class of tracers for sentinel lymph node detection, and can be extended to other radiopharmaceuticals for different targets. The statistical routine was based on three-level factorial design. Best labeling conditions were achieved. The specific activity reached was 5 MBq/{mu}g.

  8. Cleaving of TOPAS and PMMA microstructured polymer optical fibers: Core-shift and statistical quality optimization

    DEFF Research Database (Denmark)

    Stefani, Alessio; Nielsen, Kristian; Rasmussen, Henrik K.

    2012-01-01

    We fabricated an electronically controlled polymer optical fiber cleaver, which uses a razor-blade guillotine and provides independent control of fiber temperature, blade temperature, and cleaving speed. To determine the optimum cleaving conditions of microstructured polymer optical fibers (m......POFs) with hexagonal hole structures we developed a program for cleaving quality optimization, which reads in a microscope image of the fiber end-facet and determines the core-shift and the statistics of the hole diameter, hole-to-hole pitch, hole ellipticity, and direction of major ellipse axis. For 125μm in diameter...

  9. A reduced scale two loop PWR core designed with particle swarm optimization technique

    International Nuclear Information System (INIS)

    Lima Junior, Carlos A. Souza; Pereira, Claudio M.N.A; Lapa, Celso M.F.; Cunha, Joao J.; Alvim, Antonio C.M.

    2007-01-01

    Reduced scale experiments are often employed in engineering projects because they are much cheaper than real scale testing. Unfortunately, designing reduced scale thermal-hydraulic circuit or equipment, with the capability of reproducing, both accurately and simultaneously, all physical phenomena that occur in real scale and at operating conditions, is a difficult task. To solve this problem, advanced optimization techniques, such as Genetic Algorithms, have been applied. Following this research line, we have performed investigations, using the Particle Swarm Optimization (PSO) Technique, to design a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power and non accidental operating conditions. Obtained results show that the proposed methodology is a promising approach for forced flow reduced scale experiments. (author)

  10. Optimized surface-slab excited-state muffin-tin potential and surface core level shifts

    International Nuclear Information System (INIS)

    Rundgren, J.

    2003-01-01

    An optimized muffin-tin (MT) potential for surface slabs with preassigned surface core-level shifts (SCLS's) is presented. By using the MT radii as adjustable parameters the model is able to conserve the definition of the SCLS with respect to the bulk and concurrently to generate a potential that is continuous at the MT radii. The model is conceived for elastic electron scattering in a surface slab with exchange-correlation interaction described by the local density approximation. The model employs two data bases for the self-energy of the signal electron (after Hedin and Lundqvist or Sernelius). The potential model is discussed in detail with two surface structures Be(101-bar0), for which SCLS's are available, and Cu(111)p(2x2)Cs, in which the close-packed radii of the atoms are extremely different. It is considered plausible that tensor LEED based on an optimized MT potential can be used for determining SCLS's

  11. Core design optimization by integration of a fast 3-D nodal code in a heuristic search procedure

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Leege, P.F.A. de; Hoogenboom, J.E.; Quist, A.J. [Delft University of Technology, NL-2629 JB Delft (Netherlands)

    1998-07-01

    An automated design tool is being developed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, which is a 2 MWth swimming-pool type research reactor. As a black box evaluator, the 3-D nodal code SILWER, which up to now has been used only for evaluation of predetermined core designs, is integrated in the core optimization procedure. SILWER is a part of PSl's ELCOS package and features optional additional thermal-hydraulic, control rods and xenon poisoning calculations. This allows for fast and accurate evaluation of different core designs during the optimization search. Special attention is paid to handling the in- and output files for SILWER such that no adjustment of the code itself is required for its integration in the optimization programme. The optimization objective, the safety and operation constraints, as well as the optimization procedure, are discussed. (author)

  12. Core design optimization by integration of a fast 3-D nodal code in a heuristic search procedure

    International Nuclear Information System (INIS)

    Geemert, R. van; Leege, P.F.A. de; Hoogenboom, J.E.; Quist, A.J.

    1998-01-01

    An automated design tool is being developed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, which is a 2 MWth swimming-pool type research reactor. As a black box evaluator, the 3-D nodal code SILWER, which up to now has been used only for evaluation of predetermined core designs, is integrated in the core optimization procedure. SILWER is a part of PSl's ELCOS package and features optional additional thermal-hydraulic, control rods and xenon poisoning calculations. This allows for fast and accurate evaluation of different core designs during the optimization search. Special attention is paid to handling the in- and output files for SILWER such that no adjustment of the code itself is required for its integration in the optimization programme. The optimization objective, the safety and operation constraints, as well as the optimization procedure, are discussed. (author)

  13. In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

    International Nuclear Information System (INIS)

    Todt, W.H. Sr.; Playfoot, K.C.

    1988-01-01

    This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

  14. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  15. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  16. Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

    International Nuclear Information System (INIS)

    Huang Jinhua; Sun Hanhong

    1988-08-01

    Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

  17. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

    2016-04-15

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  18. A new equation of state Based on Nuclear Statistical Equilibrium for Core-Collapse Simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-09-01

    We calculate a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect on the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores.

  19. Comparative assessment of out-of-core nuclear thermionic power systems

    International Nuclear Information System (INIS)

    Estabrook, W.C.; Koenig, D.R.; Prickett, W.Z.

    1975-01-01

    The hardware selections available for fabrication of a nuclear electric propulsion stage for planetary exploration were explored. The investigation was centered around a heat-pipe-cooled, fast-spectrum nuclear reactor for an out-of-core power conversion system with sufficient detail for comparison with the in-core system studies completed previously. A survey of competing power conversion systems still indicated that the modular reliability of thermionic converters makes them the desirable choice to provide the 240-kWe end-of-life power for at least 20,000 full power hours. The electrical energy will be used to operate a number of mercury ion bombardment thrusters with a specific impulse in the range of about 4,000-5,000 seconds. (Author)

  20. Optimal ROS Signaling Is Critical for Nuclear Reprogramming

    Directory of Open Access Journals (Sweden)

    Gang Zhou

    2016-05-01

    Full Text Available Efficient nuclear reprogramming of somatic cells to pluripotency requires activation of innate immunity. Because innate immune activation triggers reactive oxygen species (ROS signaling, we sought to determine whether there was a role of ROS signaling in nuclear reprogramming. We examined ROS production during the reprogramming of doxycycline (dox-inducible mouse embryonic fibroblasts (MEFs carrying the Yamanaka factors (Oct4, Sox2, Klf4, and c-Myc [OSKM] into induced pluripotent stem cells (iPSCs. ROS generation was substantially increased with the onset of reprogramming. Depletion of ROS via antioxidants or Nox inhibitors substantially decreased reprogramming efficiency. Similarly, both knockdown and knockout of p22phox—a critical subunit of the Nox (1–4 complex—decreased reprogramming efficiency. However, excessive ROS generation using genetic and pharmacological approaches also impaired reprogramming. Overall, our data indicate that ROS signaling is activated early with nuclear reprogramming, and optimal levels of ROS signaling are essential to induce pluripotency.

  1. Contaminated Land Remediation on decommissioned nuclear facilities: an optimized approach

    International Nuclear Information System (INIS)

    Sauer, Emilie

    2016-01-01

    The site of the Monts d'Arree located in Brennilis in the area of Brittany in France is a former 70 MWe heavy water reactor. EDF is now in charge of its decommissioning. The effluent treatment facility (STE) is currently being dismantled. As the future use of the site will exclude any nuclear activity, EDF is taking site release into consideration. Therefore a land management strategy for the land and soil is needed. An optimized approach is being proposed for the STE, to the French Regulator. In France, there is no specific regulation related to contaminated land (either radiologically contaminated or chemically contaminated). The French Nuclear Safety Authority's doctrine for radioactively contaminated land is a reference approach which involves complete clean-up, removing any trace of artificial radioactivity in the ground. If technical difficulties are encountered or the quantity of radioactive waste produced is too voluminous, an optimised clean-up can be implemented. EDF has been engaged since 2008 in drawing up a common guideline with other French nuclear operators (CEA and AREVA). The operators' guideline proposed the first steps to define how to optimise nuclear waste and to carry out a cost-benefits analysis. This is in accordance with the IAEA's prescriptions. Historically, various incidents involving effluent drum spills caused radiological contamination in the building platform and the underlying soil. While conducting the decontamination works in 2004/2005, it was impossible to remove all contamination (that went deeper than expected). A large characterization campaign was carried out in order to map the contamination. For the site investigation, 34 boreholes were drilled from 2 to 5 m under the building platform and 98 samples were analyzed to search for gamma, beta and alpha emitters. With the results, the contamination was mapped using a geostatistical approach developed by Geovariances TM . Main results were: - Soils are

  2. Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

    International Nuclear Information System (INIS)

    Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

    1991-11-01

    Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

  3. Scale model study of the seismic response of a nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Dunwoody, W.E.; Rhorer, R.L.

    1983-01-01

    The use of scale models to study the dynamics of a system of graphite core blocks used in certain nuclear reactor designs is described. Scaling laws, material selecton, model instrumentation to measure collision forces, and the response of several models to simulated seismic excitation are covered. The effects of Coulomb friction between the blocks and the clearance gaps between the blocks on the system response to seismic excitation are emphasized

  4. About the application of MCNP4 code in nuclear reactor core design calculations

    International Nuclear Information System (INIS)

    Svarny, J.

    2000-01-01

    This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)

  5. In core reload design for cycle 4 of Daya Bay nuclear power station both units

    International Nuclear Information System (INIS)

    Zhang Zongyao; Liu Xudong; Xian Chunyu; Li Dongsheng; Zhang Hong; Liu Changwen; Rui Min; Wang Yingming; Zhao Ke; Zhang Hong; Xiao Min

    1998-01-01

    The basic principles and the contents of the reload design for Daya Bay nuclear power station are briefly introduced. The in core reload design results, and the comparison between the calculated values and the measured values of both units the fourth cycle are also given. The reload design results of the two units satisfy all the economic requirements and safety criteria. The experimented results shown that the predicated values are tally good with all the measurement values

  6. Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

    1985-07-01

    The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

  7. Rift Valley fever phlebovirus NSs protein core domain structure suggests molecular basis for nuclear filaments.

    Science.gov (United States)

    Barski, Michal; Brennan, Benjamin; Miller, Ona K; Potter, Jane A; Vijayakrishnan, Swetha; Bhella, David; Naismith, James H; Elliott, Richard M; Schwarz-Linek, Ulrich

    2017-09-15

    Rift Valley fever phlebovirus (RVFV) is a clinically and economically important pathogen increasingly likely to cause widespread epidemics. RVFV virulence depends on the interferon antagonist non-structural protein (NSs), which remains poorly characterized. We identified a stable core domain of RVFV NSs (residues 83-248), and solved its crystal structure, a novel all-helical fold organized into highly ordered fibrils. A hallmark of RVFV pathology is NSs filament formation in infected cell nuclei. Recombinant virus encoding the NSs core domain induced intranuclear filaments, suggesting it contains all essential determinants for nuclear translocation and filament formation. Mutations of key crystal fibril interface residues in viruses encoding full-length NSs completely abrogated intranuclear filament formation in infected cells. We propose the fibrillar arrangement of the NSs core domain in crystals reveals the molecular basis of assembly of this key virulence factor in cell nuclei. Our findings have important implications for fundamental understanding of RVFV virulence.

  8. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    Hecker, H.C.

    1984-04-01

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  9. Comment on the in-core measurement in the WWER nuclear power plant

    International Nuclear Information System (INIS)

    Krett, V.; Dach, K.; Erben, O.

    1985-01-01

    The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

  10. Optimization of emergency response to major nuclear accidents

    International Nuclear Information System (INIS)

    Papazoglou, I.A.; Christou, M.D.

    1991-01-01

    A methodology for the optimization of the short-term emergency response in the event of a nuclear accident has been developed. The method aims at an optimum combination of protective actions in the presence of a multitude of conflicting objectives and under uncertainty. Conflicting objectives arise when the minimization of the potential adverse effects of an accident and the simultaneous minimization of the associated socioeconomic impacts is attempted. Additional conflicting objectives appear whenever an emergency plan tends to decrease a particular health effect (e.g. acute deaths) while at the same time it increases another (e.g. latent deaths). The uncertainty is due to the multitude of the possible accident scenarios and their respective probability of occurrence, the stochastic variability in the weather conditions and in the variability and/or lack of knowledge in the parameters of the risk assessment models. A multiobjective optimization approach is adopted in a dynamic programming scheme. An emergency protective plan consists of defining a protective action (e.g. evacuation, sheltering) at each spatial cell around the plant. Three criteria (evaluators) are used as the objective functions of the problem, namely, acute fatalities, latent effects and socioeconomic cost. The optimization procedure defines the efficient frontier, i.e. all emergency plans that are not dominated by another in all three criteria. No value trade-offs are necessary up to this point

  11. Design and fuel management of PWR cores to optimize the once-through fuel cycle

    International Nuclear Information System (INIS)

    Fujita, E.K.; Driscoll, M.J.; Lanning, D.D.

    1978-08-01

    The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) axial power shaping by enrichment gradation in fresh fuel, (3) use of 6-batch cores with semi-annual refueling, (4) use of 6-batch cores with annual refueling, hence greater extended (approximately doubled) burnup, (5) use of radial reflector assemblies, (6) use of internally heterogeneous cores (simple seed/blanket configurations), (7) use of power/temperature coastdown at the end of life to extend burnup, (8) use of metal or diluted oxide fuel, (9) use of thorium, and (10) use of isotopically separated low sigma/sub a/ cladding material. State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications

  12. The Accumulating Data to Optimally Predict Obesity Treatment (ADOPT) Core Measures Project: Rationale and Approach.

    Science.gov (United States)

    MacLean, Paul S; Rothman, Alexander J; Nicastro, Holly L; Czajkowski, Susan M; Agurs-Collins, Tanya; Rice, Elise L; Courcoulas, Anita P; Ryan, Donna H; Bessesen, Daniel H; Loria, Catherine M

    2018-04-01

    Individual variability in response to multiple modalities of obesity treatment is well documented, yet our understanding of why some individuals respond while others do not is limited. The etiology of this variability is multifactorial; however, at present, we lack a comprehensive evidence base to identify which factors or combination of factors influence treatment response. This paper provides an overview and rationale of the Accumulating Data to Optimally Predict obesity Treatment (ADOPT) Core Measures Project, which aims to advance the understanding of individual variability in response to adult obesity treatment. This project provides an integrated model for how factors in the behavioral, biological, environmental, and psychosocial domains may influence obesity treatment responses and identify a core set of measures to be used consistently across adult weight-loss trials. This paper provides the foundation for four companion papers that describe the core measures in detail. The accumulation of data on factors across the four ADOPT domains can inform the design and delivery of effective, tailored obesity treatments. ADOPT provides a framework for how obesity researchers can collectively generate this evidence base and is a first step in an ongoing process that can be refined as the science advances. © 2018 The Obesity Society.

  13. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  14. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  15. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Mazaira, Leorlen Y.R.; Dominguez, Dany S.; Hernandez, Carlos R.G.

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  16. Applications of nuclear safety probabilistic risk assessment to nuclear security for optimized risk mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Donnelly, S.K.; Harvey, S.B. [Amec Foster Wheeler, Toronto, Ontario (Canada)

    2016-06-15

    Critical infrastructure assets such as nuclear power generating stations are potential targets for malevolent acts. Probabilistic methodologies can be applied to evaluate the real-time security risk based upon intelligence and threat levels. By employing this approach, the application of security forces and other protective measures can be optimized. Existing probabilistic safety analysis (PSA) methodologies and tools employed. in the nuclear industry can be adapted to security applications for this purpose. Existing PSA models can also be adapted and enhanced to consider total plant risk, due to nuclear safety risks as well as security risks. By creating a Probabilistic Security Model (PSM), safety and security practitioners can maximize the safety and security of the plant while minimizing the significant costs associated with security upgrades and security forces. (author)

  17. Simulation Based Optimization of Complex Monolithic Composite Structures Using Cellular Core Technology

    Science.gov (United States)

    Hickmott, Curtis W.

    Cellular core tooling is a new technology which has the capability to manufacture complex integrated monolithic composite structures. This novel tooling method utilizes thermoplastic cellular cores as inner tooling. The semi-rigid nature of the cellular cores makes them convenient for lay-up, and under autoclave temperature and pressure they soften and expand providing uniform compaction on all surfaces including internal features such as ribs and spar tubes. This process has the capability of developing fully optimized aerospace structures by reducing or eliminating assembly using fasteners or bonded joints. The technology is studied in the context of evaluating its capabilities, advantages, and limitations in developing high quality structures. The complex nature of these parts has led to development of a model using the Finite Element Analysis (FEA) software Abaqus and the plug-in COMPRO Common Component Architecture (CCA) provided by Convergent Manufacturing Technologies. This model utilizes a "virtual autoclave" technique to simulate temperature profiles, resin flow paths, and ultimately deformation from residual stress. A model has been developed simulating the temperature profile during curing of composite parts made with the cellular core technology. While modeling of composites has been performed in the past, this project will look to take this existing knowledge and apply it to this new manufacturing method capable of building more complex parts and develop a model designed specifically for building large, complex components with a high degree of accuracy. The model development has been carried out in conjunction with experimental validation. A double box beam structure was chosen for analysis to determine the effects of the technology on internal ribs and joints. Double box beams were manufactured and sectioned into T-joints for characterization. Mechanical behavior of T-joints was performed using the T-joint pull-off test and compared to traditional

  18. The Impact of the Nuclear Equation of State in Core Collapse Supernovae

    Science.gov (United States)

    Baird, M. L.; Lentz, E. J.; Hix, W. R.; Mezzacappa, A.; Messer, O. E. B.; Liebendoerfer, M.; TeraScale Supernova Initiative Collaboration

    2005-12-01

    One of the key ingredients to the core collapse supernova mechanism is the physics of matter at or near nuclear density. Included in simulations as part of the Equation of State (EOS), nuclear repulsion experienced at high densities are responsible for the bounce shock, which initially causes the outer envelope of the supernova to expand, as well as determining the structure of the newly formed proto-neutron star. Recent years have seen renewed interest in this fundamental piece of supernova physics, resulting in several promising candidate EOS parameterizations. We will present the impact of these variations in the nuclear EOS using spherically symmetric, Newtonian and General Relativistic neutrino transport simulations of stellar core collapse and bounce. This work is supported in part by SciDAC grants to the TeraScale Supernovae Initiative from the DOE Office of Science High Energy, Nuclear, and Advanced Scientific Computing Research Programs. Oak Ridge National Laboratory is managed by UT-Battelle, LLC, for U.S. Department of Energy under contract DEAC05-00OR22725

  19. Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

    International Nuclear Information System (INIS)

    Elsawi, Mohamed A.; Alnoamani, Zainab

    2015-01-01

    The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

  20. Burst shield for a pressurized nuclear-reactor core and method of operating same

    International Nuclear Information System (INIS)

    Beine, B.; Schilling, F.

    1976-01-01

    A pressurized nuclear-reactor core stands on a base up from which extends a cylindrical side wall formed of a plurality of hollow iron castings held together by circumferential and longitudinal prestressed elements. A cylindrical space between this shield and the core serves for inspection of the core and is normally filled with cast-iron segmental slabs so that if the core bursts pieces thrown out do not acquire any dangerous kinetic energy before engaging the burst shield. The top of the shield is removably secured to the side so that it can be moved out of the way periodically for removal of the filler slabs and inspection of the core. An anchor on the upper end of each longitudinal prestressing element bears against a sleeve pressing against the uppermost side element, and a nut engageable with this anchor is engageable down over the top to hold it in place, removal of this nut leaving the element prestressed in the side wall. 11 claims, 16 drawing figures

  1. Optimization of automation: III. Development of optimization method for determining automation rate in nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Min; Kim, Jong Hyun; Kim, Man Cheol; Seong, Poong Hyun

    2016-01-01

    Highlights: • We propose an appropriate automation rate that enables the best human performance. • We analyze the shortest working time considering Situation Awareness Recovery (SAR). • The optimized automation rate is estimated by integrating the automation and ostracism rate estimation methods. • The process to derive the optimized automation rate is demonstrated through case studies. - Abstract: Automation has been introduced in various industries, including the nuclear field, because it is commonly believed that automation promises greater efficiency, lower workloads, and fewer operator errors through reducing operator errors and enhancing operator and system performance. However, the excessive introduction of automation has deteriorated operator performance due to the side effects of automation, which are referred to as Out-of-the-Loop (OOTL), and this is critical issue that must be resolved. Thus, in order to determine the optimal level of automation introduction that assures the best human operator performance, a quantitative method of optimizing the automation is proposed in this paper. In order to propose the optimization method for determining appropriate automation levels that enable the best human performance, the automation rate and ostracism rate, which are estimation methods that quantitatively analyze the positive and negative effects of automation, respectively, are integrated. The integration was conducted in order to derive the shortest working time through considering the concept of situation awareness recovery (SAR), which states that the automation rate with the shortest working time assures the best human performance. The process to derive the optimized automation rate is demonstrated through an emergency operation scenario-based case study. In this case study, four types of procedures are assumed through redesigning the original emergency operating procedure according to the introduced automation and ostracism levels. Using the

  2. Neutrino-pair emission from nuclear de-excitation in core-collapse supernova simulations

    Science.gov (United States)

    Fischer, T.; Langanke, K.; Martínez-Pinedo, G.

    2013-12-01

    We study the impact of neutrino-pair production from the de-excitation of highly excited heavy nuclei on core-collapse supernova simulations, following the evolution up to several 100 ms after core bounce. Our study is based on the agile-boltztransupernova code, which features general relativistic radiation hydrodynamics and accurate three-flavor Boltzmann neutrino transport in spherical symmetry. In our simulations the nuclear de-excitation process is described in two different ways. At first we follow the approach proposed by Fuller and Meyer [Astrophys. J.AJLEEY0004-637X10.1086/170317 376, 701 (1991)], which is based on strength functions derived in the framework of the nuclear Fermi-gas model of noninteracting nucleons. Second, we parametrize the allowed and forbidden strength distributions in accordance with measurements for selected nuclear ground states. We determine the de-excitation strength by applying the Brink hypothesis and detailed balance. For both approaches, we find that nuclear de-excitation has no effect on the supernova dynamics. However, we find that nuclear de-excitation is the leading source for the production of electron antineutrinos as well as heavy-lepton-flavor (anti)neutrinos during the collapse phase. At sufficiently high densities, the associated neutrino spectra are influenced by interactions with the surrounding matter, making proper simulations of neutrino transport important for the determination of the neutrino-energy loss rate. We find that, even including nuclear de-excitations, the energy loss during the collapse phase is overwhelmingly dominated by electron neutrinos produced by electron capture.

  3. Application of optimization methods for nuclear energy system performance assessment by the MESSAGE software

    International Nuclear Information System (INIS)

    Andrianov, A.A.; Kuptsov, I.S.; Utyanskaya, T.V.

    2016-01-01

    This paper defines the multi-objective optimization and uncertainty treatment modules for the IAEA energy planning software MESSAGE intended for multi-objective optimization and sustainability assessments of innovative nuclear energy systems with account of uncertainty [ru

  4. The maintenance optimization of structural components in nuclear power plants

    International Nuclear Information System (INIS)

    Bryla, P.; Ardorino, F.; Aufort, P.; Jacquot, J.P.; Magne, L.; Pitner, P.; Verite, B.; Villain, B.; Monnier, B.

    1997-10-01

    An optimization process, called 'OMF-Structures', is developed by Electricite de France (EDF) in order to extend the current 'OMF' Reliability Centered Maintenance to piping structural components. The Auxiliary Feedwater System of a 900 MW French nuclear plant has been studied in order to lay the foundations of the method. This paper presents the currently proposed principles of the process. The principles of the OMF-Structures process include 'Risk-Based Inspection' concepts within an RCM process. Two main phases are identified: The purpose of the first phase is to select the risk-significant failure modes and associated elements. This phase consists of two major steps: potential consequences evaluation and reliability performance evaluation. The second phase consists of the definition of preventive maintenance programs for piping elements that are associated with risk-significant failure modes. (author)

  5. Nuclear steam generator tube to tubesheet joint optimization

    International Nuclear Information System (INIS)

    McGregor, Rod

    1999-01-01

    Industry-wide problems with Stress Corrosion Cracking in the Nuclear Steam Generator tube-to-tubesheet joint have led to costly repairs, plugging, and replacement of entire vessels. To improve corrosion resistance, new and replacement Steam Generator developments typically employ the hydraulic tube expansion process (full depth) to minimize tensile residual stresses and cold work at the critical transition zone between the expanded and unexpanded tube. These variables have undergone detailed study using specialized X-ray diffraction and analytical techniques. Responding to increased demands from Nuclear Steam Generator operators and manufacturers to credit the leak-tightness and strength contributions of the hydraulic expansion, various experimental tasks with complimentary analytical modelling were applied to improve understanding and control of tube to hole contact pressure. With careful consideration to residual stress impact, design for strength/leak tightness optimization addresses: Experimentally determined minimum contact pressure levels necessary to preclude incipient leakage into the tube/hole interface. The degradation of contact pressure at surrounding expansions caused by the sequential expansion process. The transient and permanent contact pressure variation associated with tubesheet hole dilation during Steam Generator operation. An experimental/analytical simulation has been developed to reproduce cyclic Steam Generator operating strains on the tubesheet and expanded joint. Leak tightness and pullout tests were performed during and following simulated Steam Generator operating transients. The overall development has provided a comprehensive understanding of the fabrication and in-service mechanics of hydraulically expanded joints. Based on this, the hydraulic expansion process can be optimized with respect to critical residual stress/cold work and the strength/leakage barrier criteria. (author)

  6. System identification using Nuclear Norm & Tabu Search optimization

    Science.gov (United States)

    Ahmed, Asif A.; Schoen, Marco P.; Bosworth, Ken W.

    2018-01-01

    In recent years, subspace System Identification (SI) algorithms have seen increased research, stemming from advanced minimization methods being applied to the Nuclear Norm (NN) approach in system identification. These minimization algorithms are based on hard computing methodologies. To the authors’ knowledge, as of now, there has been no work reported that utilizes soft computing algorithms to address the minimization problem within the nuclear norm SI framework. A linear, time-invariant, discrete time system is used in this work as the basic model for characterizing a dynamical system to be identified. The main objective is to extract a mathematical model from collected experimental input-output data. Hankel matrices are constructed from experimental data, and the extended observability matrix is employed to define an estimated output of the system. This estimated output and the actual - measured - output are utilized to construct a minimization problem. An embedded rank measure assures minimum state realization outcomes. Current NN-SI algorithms employ hard computing algorithms for minimization. In this work, we propose a simple Tabu Search (TS) algorithm for minimization. TS algorithm based SI is compared with the iterative Alternating Direction Method of Multipliers (ADMM) line search optimization based NN-SI. For comparison, several different benchmark system identification problems are solved by both approaches. Results show improved performance of the proposed SI-TS algorithm compared to the NN-SI ADMM algorithm.

  7. A Novel Method To On-Line Monitor Reactor Nuclear Power And In-Core Thermal Environments

    International Nuclear Information System (INIS)

    Liu, Hanying; Miller, Don W.; Li, Dongxu; Radcliff, Thomas D.

    2002-01-01

    For current nuclear power plants, nuclear power can not be directly measured and in-core fuel thermal environments can not be monitored due to the unavailability of an appropriate measurement technology and the inaccessibility of the fuel. If the nuclear deposited power and the in-core thermal conditions (i.e. fuel or coolant temperature and heat transfer coefficient) can be monitored in-situ, then it would play a valuable and critical role in increasing nuclear power, predicting abnormal reactor operation, improving core physical models and reducing core thermal margin so as to implement higher fuel burn-up. Furthermore, the management of core thermal margin and fuel operation may be easier during reactor operation, post-accident or spent fuel storage. On the other hand, for some advanced Generation IV reactors, the sealed and long-lived reactor core design challenges traditional measurement techniques while conventional ex-core detectors and current in-core detectors can not monitor details of the in-core fuel conditions. A method is introduced in this paper that responds to the challenge to measure nuclear power and to monitor the in-core thermal environments, for example, local fuel pin or coolant heat convection coefficient and temperature. In summary, the method, which has been designed for online in-core measurement and surveillance, will be beneficial to advanced plant safety, efficiency and economics by decreasing thermal margin or increasing nuclear power. The method was originally developed for a constant temperature power sensor (CTPS). The CTPS is undergoing design and development for an advanced reactor core to measure in-core nuclear power in measurement mode and to monitor thermal environments in compensation mode. The sensor dynamics was analyzed in compensation mode to determine the environmental temperature and the heat transfer coefficient. Previous research demonstrated that a first order dynamic model is not sufficient to simulate sensor

  8. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Design and performance of reference cores. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely of tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. The effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  9. An improved heat transfer configuration for a solid-core nuclear thermal rocket engine

    International Nuclear Information System (INIS)

    Clark, J.S.; Walton, J.T.; Mcguire, M.L.

    1992-07-01

    Interrupted flow, impingement cooling, and axial power distribution are employed to enhance the heat-transfer configuration of a solid-core nuclear thermal rocket engine. Impingement cooling is introduced to increase the local heat-transfer coefficients between the reactor material and the coolants. Increased fuel loading is used at the inlet end of the reactor to enhance heat-transfer capability where the temperature differences are the greatest. A thermal-hydraulics computer program for an unfueled NERVA reactor core is employed to analyze the proposed configuration with attention given to uniform fuel loading, number of channels through the impingement wafers, fuel-element length, mass-flow rate, and wafer gap. The impingement wafer concept (IWC) is shown to have heat-transfer characteristics that are better than those of the NERVA-derived reactor at 2500 K. The IWC concept is argued to be an effective heat-transfer configuration for solid-core nuclear thermal rocket engines. 11 refs

  10. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  11. Simple heuristics: A bridge between manual core design and automated optimization methods

    International Nuclear Information System (INIS)

    White, J.R.; Delmolino, P.M.

    1993-01-01

    The primary function of RESCUE is to serve as an aid in the analysis and identification of feasible loading patterns for LWR reload cores. The unique feature of RESCUE is that its physics model is based on some recent advances in generalized perturbation theory (GPT) methods. The high order GPT techniques offer the accuracy, computational efficiency, and flexibility needed for the implementation of a full range of capabilities within a set of compatible interactive (manual and semi-automated) and automated design tools. The basic design philosophy and current features within RESCUE are reviewed, and the new semi-automated capability is highlighted. The online advisor facility appears quite promising and it provides a natural bridge between the traditional trial-and-error manual process and the recent progress towards fully automated optimization sequences. (orig.)

  12. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

    1991-12-31

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

  13. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

    1990-01-01

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

  14. Accumulating Data to Optimally Predict Obesity Treatment (ADOPT) Core Measures: Behavioral Domain.

    Science.gov (United States)

    Lytle, Leslie A; Nicastro, Holly L; Roberts, Susan B; Evans, Mary; Jakicic, John M; Laposky, Aaron D; Loria, Catherine M

    2018-04-01

    The ability to identify and measure behaviors that are related to weight loss and the prevention of weight regain is crucial to understanding the variability in response to obesity treatment and the development of tailored treatments. The overarching goal of the Accumulating Data to Optimally Predict obesity Treatment (ADOPT) Core Measures Project is to provide obesity researchers with guidance on a set of constructs and measures that are related to weight control and that span and integrate obesity-related behavioral, biological, environmental, and psychosocial domains. This article describes how the behavioral domain subgroup identified the initial list of high-priority constructs and measures to be included, and it describes practical considerations for assessing the following four behavioral areas: eating, activity, sleep, and self-monitoring of weight. Challenges and considerations for advancing the science related to weight loss and maintenance behaviors are also discussed. Assessing a set of core behavioral measures in combination with those from other ADOPT domains is critical to improve our understanding of individual variability in response to adult obesity treatment. The selection of behavioral measures is based on the current science, although there continues to be much work needed in this field. © 2018 The Obesity Society.

  15. In core instrumentation for online nuclear heating measurements of material testing reactor

    International Nuclear Information System (INIS)

    Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

    2010-01-01

    The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

  16. Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

    International Nuclear Information System (INIS)

    Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

    1985-11-01

    A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

  17. Multi-physics and multi-objective design of heterogeneous SFR core: development of an optimization method under uncertainty

    International Nuclear Information System (INIS)

    Ammar, Karim

    2014-01-01

    Since Phenix shutting down in 2010, CEA does not have Sodium Fast Reactor (SFR) in operating condition. According to global energetic challenge and fast reactor abilities, CEA launched a program of industrial demonstrator called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a reactor with electric power capacity equal to 600 MW. Objective of the prototype is, in first to be a response to environmental constraints, in second demonstrates the industrial viability of SFR reactor. The goal is to have a safety level at least equal to 3. generation reactors. ASTRID design integrates Fukushima feedback; Waste reprocessing (with minor actinide transmutation) and it linked industry. Installation safety is the priority. In all cases, no radionuclide should be released into environment. To achieve this objective, it is imperative to predict the impact of uncertainty sources on reactor behaviour. In this context, this thesis aims to develop new optimization methods for SFR cores. The goal is to improve the robustness and reliability of reactors in response to existing uncertainties. We will use ASTRID core as reference to estimate interest of new methods and tools developed. The impact of multi-Physics uncertainties in the calculation of the core performance and the use of optimization methods introduce new problems: How to optimize 'complex' cores (i.e. associated with design spaces of high dimensions with more than 20 variable parameters), taking into account the uncertainties? What is uncertainties behaviour for optimization core compare to reference core? Taking into account uncertainties, optimization core are they still competitive? Optimizations improvements are higher than uncertainty margins? The thesis helps to develop and implement methods necessary to take into account uncertainties in the new generation of simulation tools. Statistical methods to ensure consistency of complex multi-Physics simulation results are also

  18. Spent nuclear fuel application of CORE reg-sign systems engineering software

    International Nuclear Information System (INIS)

    Grimm, R.J.

    1996-01-01

    The Department of Energy (DOE) has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF Program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF Program. The sheer size and complexity of the SNF Program, however, has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off-the-shelf PC based software package, to assist the DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

  19. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    International Nuclear Information System (INIS)

    Prabhu Gaunkar, N.; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C.; Bulu, I.; Ganesan, K.; Song, Y. Q.

    2015-01-01

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors

  20. Spent nuclear fuel application of CORE reg-sign systems engineering software

    International Nuclear Information System (INIS)

    Grimm, R.J.

    1996-01-01

    The DOE has adopted a systems engineering approach for the successful completion of the Spent Nuclear Fuel (SNF) Program mission. The DOE has utilized systems engineering principles to develop the SNF program guidance documents and has held several systems engineering workshops to develop the functional hierarchies of both the programmatic and technical side of the SNF program. The sheer size and complexity of the SNF program has led to problems that the Westinghouse Savannah River Company (WSRC) is working to manage through the use of systems engineering software. WSRC began using CORE reg-sign, an off the shelf PC based software package, to assist DOE in management of the SNF program. This paper details the successful use of the CORE reg-sign systems engineering software to date and the proposed future activities

  1. Control and balance of nuclear matters used for core fabrication of Super Phenix

    International Nuclear Information System (INIS)

    Beche, M.; Guillet, H.; Heyraud, H.; Levrard, J.; Pajot, J.

    1987-05-01

    The fabrication of the core of the fast breeder reactor set up at Creys Malville ended in March 1984. It started in 1978 and it required, for the fabrication of the 410 assemblies, the utilization of 7438 kg of plutonium. To satisfy national and international regulations, DPFER/SFER has used a methodology to follow and to control the movements of the nuclear materials. These controls are achieved by physical methods, chemical methods and empiric methods. Euratom has conducted a succession of inspections during the 5.5 years of that campaign. The inventory difference, in the fabrication of that core, represents about 0.1% of the total mass of the plutonium handled [fr

  2. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases

  3. Impact of correlations between core configurations for the evaluation of nuclear data uncertainty propagation for reactivity

    International Nuclear Information System (INIS)

    Frosio, T.; Bonaccorsi, T.; Blaise, P.

    2017-01-01

    The precise estimation of Pearson correlations, also called 'representativity' coefficients, between core configurations is a fundamental quantity for properly assessing the nuclear data (ND) uncertainties propagation on integral parameters such as k-eff, power distributions, or reactivity coefficients. In this paper, a traditional adjoint method is used to propagate ND uncertainty on reactivity and reactivity coefficients and estimate correlations between different states of the core. We show that neglecting those correlations induces a loss of information in the final uncertainty. We also show that using approximate values of Pearson does not lead to an important error of the model. This calculation is made for reactivity at the beginning of life and can be extended to other parameters during depletion calculations. (authors)

  4. Analysis of ringing effects due to magnetic core materials in pulsed nuclear magnetic resonance circuits

    Energy Technology Data Exchange (ETDEWEB)

    Prabhu Gaunkar, N., E-mail: neelampg@iastate.edu; Bouda, N. R. Y.; Nlebedim, I. C.; Hadimani, R. L.; Mina, M.; Jiles, D. C. [Department of Electrical and Computer Engineering, Iowa State University, Ames, Iowa 50011 (United States); Bulu, I.; Ganesan, K.; Song, Y. Q. [Schlumberger-Doll Research, Cambridge, Massachusetts 02139 (United States)

    2015-05-07

    This work presents investigations and detailed analysis of ringing in a non-resonant pulsed nuclear magnetic resonance (NMR) circuit. Ringing is a commonly observed phenomenon in high power switching circuits. The oscillations described as ringing impede measurements in pulsed NMR systems. It is therefore desirable that those oscillations decay fast. It is often assumed that one of the causes behind ringing is the role of the magnetic core used in the antenna (acting as an inductive load). We will demonstrate that an LRC subcircuit is also set-up due to the inductive load and needs to be considered due to its parasitic effects. It is observed that the parasitics associated with the inductive load become important at certain frequencies. The output response can be related to the response of an under-damped circuit and to the magnetic core material. This research work demonstrates and discusses ways of controlling ringing by considering interrelationships between different contributing factors.

  5. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

  6. The effect of uncertainties in nuclear reactor plant-specific failure data on core damage frequency

    International Nuclear Information System (INIS)

    Martz, H.F.

    1995-05-01

    It is sometimes the case in PRA applications that reported plant-specific failure data are, in fact, only estimates which are uncertain. Even for detailed plant-specific data, the reported exposure time or number of demands is often only an estimate of the actual exposure time or number of demands. Likewise the reported number of failure events or incidents is sometimes also uncertain because incident or malfunction reports may be ambiguous. In this report we determine the corresponding uncertainty in core damage frequency which can b attributed to such uncertainties in plant-specific data using a simple but typical nuclear power reactor example

  7. Interaction between core analysis methodology and nuclear design: some PWR examples

    International Nuclear Information System (INIS)

    Rothleder, B.M.; Eich, W.J.

    1982-01-01

    The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

  8. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  9. Analysis of the documents about the core envelopment of nuclear reactor at the Laguna Verde U-1 power plant

    International Nuclear Information System (INIS)

    Zamora R, L.; Medina F, A.

    1999-01-01

    The degradation of internal components at BWR type reactors is an important subject to consider in the performance availability of the power plant. The Wuergassen nuclear reactor license was confiscated due to the presence of cracking in the core envelopment. In consequence it is necessary carrying out a detailed study with the purpose to avoid these problems in the future. This report presents a review and analysis of documents and technical information referring to the core envelopment of a BWR/5/6 and the Laguna Verde Unit 1 nuclear reactor in Mexico. In this document are presented design data, documents about fabrication processes, and manufacturing of core envelopment. (Author)

  10. Optimal design for crosstalk analysis in 12-core 5-LP mode homogeneous multicore fiber for different lattice structure

    Science.gov (United States)

    Kumar, Dablu; Ranjan, Rakesh

    2018-03-01

    12-Core 5-LP mode homogeneous multicore fibers have been proposed for analysis of inter-core crosstalk and dispersion, with four different lattice structures (circular, 2-ring, square lattice, and triangular lattice) having cladding diameter of 200 μm and a fixed cladding thickness of 35 μm. The core-to-core crosstalk impact has been studied numerically with respect to bending radius, core pitch, transmission distance, wavelength, and core diameter for all 5-LP modes. In anticipation of further reduction in crosstalk levels, the trench-assisted cores have been incorporated for all respective designs. Ultra-low crosstalk (-138 dB/100 km) has been achieved through the triangular lattice arrangement, with trench depth Δ2 = -1.40% for fundamental (LP01) mode. It has been noted that the impact of mode polarization on crosstalk behavior is minor, with difference in crosstalk levels between two polarized spatial modes as ≤0.2 dB. Moreover, the optimized cladding diameter has been obtained for all 5-LP modes for a target value of crosstalk of -50 dB/100 km, with all the core arrangements. The dispersion characteristic has also been analyzed with respect to wavelength, which is nearly 2.5 ps/nm km at operating wavelength 1550 nm. The relative core multiplicity factor (RCMF) for the proposed design is obtained as 64.

  11. Research on application of complex-genetic algorithm in nuclear component optimal design

    International Nuclear Information System (INIS)

    He Shijing; Yan Changqi; Wang Jianjun; Wang Meng

    2010-01-01

    Complex algorithm is one of the most commonly used methods in the mechanical design optimization, such as the optimization of nuclear component. An improved method,complex-genetic algorithm(CGA), is developed based on traditional complex algorithm(TCA), in which the disadvantages of TCA have been overcome. An optimal calculation,which represents the pressurizer, is carried out in order to analyze the optimization capability of CGA. The results show that CGA has better optimizing performance than TCA. (authors)

  12. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  13. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    International Nuclear Information System (INIS)

    Friess, Friederike Renate

    2017-01-01

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  14. Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Friess, Friederike Renate

    2017-07-12

    According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

  15. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 1. ZPPR analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-05-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess of Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. The data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS and ZPPR fast reactor cores applying JNC core calculation code CITATION. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the first flight collision probability method and subgroup approach. Especially a converting program was written to transmit the prepared effective cross sections to JNC standard PDS files. Then the CITATION code was applied for 3-D XYZ neutronics calculations of BFS and ZPPR JUPITER experiments series cores. The effects of nuclear data library have been studied by comparing the former results based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for ZPPR-9, ZPPR-13A and ZPPR-17A cores are presented. The calculated correction factor in all cases was less than 1.0%. So the uncertainty in C value caused by possible errors in calculation of these corrections is expected to be less than 0.3% in case of ZPPR-13A and ZPPR-17A cores, and rather less for ZPPR-9 core. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC calculation route revealed a large enough discrepancy in k-eff for ZPPR-9 (about 0.6%) and ZPPR-17A (about 0.5%) cores. For BFS-62-1 and BFS-62-2 cores such analysis is in progress. Stretch cell models for both BFS cores were formed and cell calculations using FFCP code have started. Some results of cell calculations are presented. (author)

  16. Optimizing Liquid Effluent Monitoring at a Large Nuclear Complex

    International Nuclear Information System (INIS)

    Chou, Charissa J.; Johnson, V.G.; Barnett, Brent B.; Olson, Phillip M.

    2003-01-01

    Monitoring data for a centralized effluent treatment and disposal facility at the Hanford Site, a defense nuclear complex undergoing cleanup and decommissioning in southeast Washington State, was evaluated to optimize liquid effluent monitoring efficiency. Wastewater from several facilities is collected and discharged to the ground at a common disposal site. The discharged water infiltrates through 60 m of soil column to the groundwater, which eventually flows into the Columbia River, the second largest river in the contiguous United States. Protection of this important natural resource is the major objective of both cleanup and groundwater and effluent monitoring activities at the Hanford Site. Four years of effluent data were evaluated for this study. More frequent sampling was conducted during the first year of operation to assess temporal variability in analyte concentrations, to determine operational factors contributing to waste stream variability and to assess the probability of exceeding permit limits. Subsequently, the study was updated which included evaluation of the sampling and analysis regime. It was concluded that the probability of exceeding permit limits was one in a million under normal operating conditions, sampling frequency could be reduced, and several analytes could be eliminated, while indicators could be substituted for more expensive analyses. Findings were used by the state regulatory agency to modify monitoring requirements for a new discharge permit. The primary focus of this paper is on the statistical approaches and rationale that led to the successful permit modification and to a more cost-effective effluent monitoring program

  17. Optimizing liquid effluent monitoring at a large nuclear complex.

    Science.gov (United States)

    Chou, Charissa J; Barnett, D Brent; Johnson, Vernon G; Olson, Phil M

    2003-12-01

    Effluent monitoring typically requires a large number of analytes and samples during the initial or startup phase of a facility. Once a baseline is established, the analyte list and sampling frequency may be reduced. Although there is a large body of literature relevant to the initial design, few, if any, published papers exist on updating established effluent monitoring programs. This paper statistically evaluates four years of baseline data to optimize the liquid effluent monitoring efficiency of a centralized waste treatment and disposal facility at a large defense nuclear complex. Specific objectives were to: (1) assess temporal variability in analyte concentrations, (2) determine operational factors contributing to waste stream variability, (3) assess the probability of exceeding permit limits, and (4) streamline the sampling and analysis regime. Results indicated that the probability of exceeding permit limits was one in a million under normal facility operating conditions, sampling frequency could be reduced, and several analytes could be eliminated. Furthermore, indicators such as gross alpha and gross beta measurements could be used in lieu of more expensive specific isotopic analyses (radium, cesium-137, and strontium-90) for routine monitoring. Study results were used by the state regulatory agency to modify monitoring requirements for a new discharge permit, resulting in an annual cost savings of US dollars 223,000. This case study demonstrates that statistical evaluation of effluent contaminant variability coupled with process knowledge can help plant managers and regulators streamline analyte lists and sampling frequencies based on detection history and environmental risk.

  18. Potential utilities of optimal estimation and control in nuclear power plants

    International Nuclear Information System (INIS)

    Tylee, J.L.; Purviance, J.E.

    1983-01-01

    Optimal estimation and control theories offer the potential for more precise control and diagnosis of nuclear power plants. The important element of these theories is that a mathematical plant model is used in conjunction with the actual plant data to optimize some performance criteria. These criteria involve important plant variables and incorporate a sense of the desired plant performance. Several applications of optimal estimation and control to nuclear systems are discussed

  19. The Use of Hidden Markov Models for Anomaly Detection in Nuclear Core Condition Monitoring

    Science.gov (United States)

    Stephen, Bruce; West, Graeme M.; Galloway, Stuart; McArthur, Stephen D. J.; McDonald, James R.; Towle, Dave

    2009-04-01

    Unplanned outages can be especially costly for generation companies operating nuclear facilities. Early detection of deviations from expected performance through condition monitoring can allow a more proactive and managed approach to dealing with ageing plant. This paper proposes an anomaly detection framework incorporating the use of the Hidden Markov Model (HMM) to support the analysis of nuclear reactor core condition monitoring data. Fuel Grab Load Trace (FGLT) data gathered within the UK during routine refueling operations has been seen to provide information relating to the condition of the graphite bricks that comprise the core. Although manual analysis of this data is time consuming and requires considerable expertise, this paper demonstrates how techniques such as the HMM can provide analysis support by providing a benchmark model of expected behavior against which future refueling events may be compared. The presence of anomalous behavior in candidate traces is inferred through the underlying statistical foundation of the HMM which gives an observation likelihood averaged along the length of the input sequence. Using this likelihood measure, the engineer can be alerted to anomalous behaviour, indicating data which might require further detailed examination. It is proposed that this data analysis technique is used in conjunction with other intelligent analysis techniques currently employed to analyse FGLT to provide a greater confidence measure in detecting anomalous behaviour from FGLT data.

  20. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Energy Technology Data Exchange (ETDEWEB)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki [Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Department of Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan) and Advanced Research Institute for Science and Engineering, Waseda University, 3-4-1 Okubo, Shinjuku, Tokyo 169-8555 (Japan); Numazu College of Technology, Ooka 3600, Numazu, Shizuoka 410-8501 (Japan); Faculty of Science and Technology, Tokyo University of Science, Yamazaki 2641, Noda, Chiba 278-8510 (Japan)

    2012-11-12

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to {approx} 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  1. Roles of nuclear weak rates on the evolution of degenerate cores in stars

    Directory of Open Access Journals (Sweden)

    Suzuki Toshio

    2017-01-01

    Full Text Available Electron-capture and β-decay rates in stellar environments are evaluated with the use of new shell-model Hamiltonians for sd-shell and pf-shell nuclei as well as for nuclei belonging to the island of inversion. Important role of the nuclear weak rates on the final evolution of stellar degenerate cores is presented. The weak interaction rates for sd-shell nuclei are calculated to study nuclear Urca processes in O-Ne-Mg cores of stars with 8-10 M⊙ (solar mass and their effects on the final fate of the stars. Nucleosynthesis of iron-group elements in Type Ia supernova explosions are studied with the weak rates for pf-shell nuclei. The problem of the neutron-rich iron-group isotope over-production compared to the solar abundances is shown to be nearly solved with the use of the new rates and explosion model of slow defraglation with delayed detonation. Evaluation of the weak rates is extended to the island of inversion and the region of neutron-rich nuclei near 78Ni, where two major shells contribute to their configurations.

  2. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    International Nuclear Information System (INIS)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-01-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ∼ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  3. A new baryonic equation of state at sub-nuclear densities for core-collapse simulations

    Science.gov (United States)

    Furusawa, Shun; Yamada, Shoichi; Sumiyoshi, Kohsuke; Suzuki, Hideyuki

    2012-11-01

    We construct a new equation of state for baryons at sub-nuclear densities for the use in core-collapse simulations of massive stars. The formulation is based on the nuclear statistical equilibrium description and the liquid drop approximation of nuclei. The model free energy to minimize is calculated by using relativistic mean field theory for nucleons and the mass formula for nuclei with atomic number up to ~ 1000. We have also taken into account the pasta phase. We find that the free energy and other thermodynamical quantities are not very different from those given in the standard EOSs that adopt the single nucleus approximation. On the other hand, the average mass is systematically different, which may have an important effect to the rates of electron captures and coherent neutrino scatterings on nuclei in supernova cores. It is also interesting that the root mean square of the mass number is not very different from the average mass number, since the former is important for the evaluation of coherent scattering rates on nuclei but has been unavailable so far.

  4. Diffusion induced nuclear reactions in metals: a possible source of heat in the core

    International Nuclear Information System (INIS)

    Hamza, V.M.; Iyer, S.S.S.

    1989-01-01

    It has recently been proposed that diffusion of light nuclei in metals can give rise to unusual electrical charge distributions in their lattice structures, inducing thereby certain nuclear reactions that are otherwise uncommon. In the light of these results we advance the hypothesis that such nuclear reactions take place in the metal rich core of the earth, based on following observations: 1 - The solubility of hydrogen in metals is relatively high compared to that in silicates. 2 - Studies of rare gas samples in intraplate volcanos and diamonds show that 3 He/ He ratio increases with depth in the mantle. 3 - There are indications that He is positively correlated with enrichment of metals in lavas. We propose that hydrogen incorporated into metallic phases at the time of planetary accretion was carried to the core by downward migration of metal rich melts during the early states of proto-earth. Preliminary estimates suggest that cold fusion reactions can give rise to an average rate of heat generation of 8.2x10 12 W and may thus serve as a supplementary source of energy for the geomagnetic dynamo. (author)

  5. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  6. Optimization on fixed low latency implementation of the GBT core in FPGA

    Science.gov (United States)

    Chen, K.; Chen, H.; Wu, W.; Xu, H.; Yao, L.

    2017-07-01

    In the upgrade of ATLAS experiment [1], the front-end electronics components are subjected to a large radiation background. Meanwhile high speed optical links are required for the data transmission between the on-detector and off-detector electronics. The GBT architecture and the Versatile Link (VL) project are designed by CERN to support the 4.8 Gbps line rate bidirectional high-speed data transmission which is called GBT link [2]. In the ATLAS upgrade, besides the link with on-detector, the GBT link is also used between different off-detector systems. The GBTX ASIC is designed for the on-detector front-end, correspondingly for the off-detector electronics, the GBT architecture is implemented in Field Programmable Gate Arrays (FPGA). CERN launches the GBT-FPGA project to provide examples in different types of FPGA [3]. In the ATLAS upgrade framework, the Front-End LInk eXchange (FELIX) system [4, 5] is used to interface the front-end electronics of several ATLAS subsystems. The GBT link is used between them, to transfer the detector data and the timing, trigger, control and monitoring information. The trigger signal distributed in the down-link from FELIX to the front-end requires a fixed and low latency. In this paper, several optimizations on the GBT-FPGA IP core are introduced, to achieve a lower fixed latency. For FELIX, a common firmware will be used to interface different front-ends with support of both GBT modes: the forward error correction mode and the wide mode. The modified GBT-FPGA core has the ability to switch between the GBT modes without FPGA reprogramming. The system clock distribution of the multi-channel FELIX firmware is also discussed in this paper.

  7. Optimization on fixed low latency implementation of the GBT core in FPGA

    International Nuclear Information System (INIS)

    Chen, K.; Chen, H.; Wu, W.; Xu, H.; Yao, L.

    2017-01-01

    In the upgrade of ATLAS experiment [1], the front-end electronics components are subjected to a large radiation background. Meanwhile high speed optical links are required for the data transmission between the on-detector and off-detector electronics. The GBT architecture and the Versatile Link (VL) project are designed by CERN to support the 4.8 Gbps line rate bidirectional high-speed data transmission which is called GBT link [2]. In the ATLAS upgrade, besides the link with on-detector, the GBT link is also used between different off-detector systems. The GBTX ASIC is designed for the on-detector front-end, correspondingly for the off-detector electronics, the GBT architecture is implemented in Field Programmable Gate Arrays (FPGA). CERN launches the GBT-FPGA project to provide examples in different types of FPGA [3]. In the ATLAS upgrade framework, the Front-End LInk eXchange (FELIX) system [4, 5] is used to interface the front-end electronics of several ATLAS subsystems. The GBT link is used between them, to transfer the detector data and the timing, trigger, control and monitoring information. The trigger signal distributed in the down-link from FELIX to the front-end requires a fixed and low latency. In this paper, several optimizations on the GBT-FPGA IP core are introduced, to achieve a lower fixed latency. For FELIX, a common firmware will be used to interface different front-ends with support of both GBT modes: the forward error correction mode and the wide mode. The modified GBT-FPGA core has the ability to switch between the GBT modes without FPGA reprogramming. The system clock distribution of the multi-channel FELIX firmware is also discussed in this paper.

  8. Nuclear reactor core having nuclear fuel and composite burnable absorber arranged for power peaking and moderator temperature coefficient control

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1991-01-01

    This patent describes a nuclear reactor core. It comprises a first group of fuel rods containing fissionable material and being free of burnable absorber material; and a second group of fuel rods containing fissionable material and first and second burnable absorber material; the first burnable absorber material being a boron-bearing material which does not contain erbium and the second burnable absorber material being an erbium material; the first and second burnable absorber materials being in the form of an outer coating on the fissionable material, the outer coating being composed of an inner layer of one of the boron-bearing material which does not contain erbium and the erbium material and an outer layer of the other of the boron-bearing material which does not contain erbium and the erbium material

  9. Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

    International Nuclear Information System (INIS)

    Mitchell, Jonathan; Fordham, Edmund J.

    2014-01-01

    Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general

  10. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  11. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  12. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  13. Application of mapping crossover genetic algorithm in nuclear power equipment optimization design

    International Nuclear Information System (INIS)

    Li Guijiang; Yan Changqi; Wang Jianjun; Liu Chengyang

    2013-01-01

    Genetic algorithm (GA) has been widely applied in nuclear engineering. An improved method, named the mapping crossover genetic algorithm (MCGA), was developed aiming at improving the shortcomings of traditional genetic algorithm (TGA). The optimal results of benchmark problems show that MCGA has better optimizing performance than TGA. MCGA was applied to the reactor coolant pump optimization design. (authors)

  14. Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

    International Nuclear Information System (INIS)

    Sevon, T.

    2005-11-01

    In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

  15. Introduction of virtual detectors for core monitoring system of korean standard nuclear power plant

    International Nuclear Information System (INIS)

    Eun, Ki Lee.; Yong, Hee Kim.; Jybe, Ho Cha.; Moon, Ghu Park.

    2000-01-01

    A novel algorithm known as the virtual detector method (VDM) is introduced to reconstruct the axial power shape (APS) for the on-line core monitoring system of the Korean Standard Nuclear Power Plant (KSNP). A pure statistical method (SM) is also introduced and the results are compared with the currently implemented five-mode Fourier fitting method (FFM). VDM adopts nine virtual detector informations coupled with a regression model based on the Alternating Conditional Expectation (ACE) algorithm. VDM uses Fourier fitting with the information of nine virtual detectors expanded from the currently implemented FFM, which uses five-level detector information. By introducing virtual detectors, we can increase the number of axial detectors, and thus expect the computational errors of APS to be reduced. The two methods (SM and VDM) are applied to in-core mapping data from six cycles of Yong Gwang nuclear power plant Units 3 and 4. For ∼ 3500 cases of APSs extracted from a cycle of operation which is simulated by a three-dimensional nodal code, the accuracy of the three methods (SM, VDM, FFM) is compared. The average root mean square (RMS) error and average of axial peaking error of SM and VDM resulted in reduction of more than 50 % and 70 %, respectively, relative to FFM. VDM and SM also show more realistic axial profiles and predict more accurate axial peaking than FFM. These improvements can contribute to a larger thermal margin. SM shows the most accurate results for all cases. VDM can almost obtain the same results as SM, and using far fewer computation steps. VDM can be a useful tool for precisely reconstructing axial power shapes in a core monitoring system. (authors)

  16. An automated optimization of core fuel loading pattern for pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Renji

    1988-11-01

    An optimum method was adopted to search for an optimum fuel loading pattern in pressurized water reactors. A radial power peak factor was chosen as the objective function of the optimum loading. The direct search method with shuffling rules is used to find optimum solution. The search for an optimum loading pattern with the smallest radial power peak by exchanging fuel assemblies was made. The search process is divided into two steps. In the first step fresh fuels or high reactivity fuels are arranged which are placed in core interior to have a reasonable fuel loading pattern. To further reduce the radial power peak factor, the second step will be necessary to rearrange the exposed lower reactivity fuel around the assemblies which has the radial power peak. In optimum process 1.5 group coarse mesh diffusion theory or two group nodal Green function diffusion theory is utilized to calculate the two dimensional power distribution after each shuffle. Also, above two methods are combinatively utilized to calculate the state quantity. It is not only true to save CPU time, but also can obtian exact results. Besides above function, the code MSOFEL is used to search critical boron concentration and to predict burn-up. The code has been written with FORTRAN-4. The optimum loading pattern was chosen for OCONEE and QINSHAN nuclear power plants as reference examples. The validity and feasibility of MSOFEL was demonstrated

  17. Performance Improvement of the Core Protection Calculator System (CPCS) by Introducing Optimal Function Sets

    International Nuclear Information System (INIS)

    Won, Byung Hee; Kim, Kyung O; Kim, Jong Kyung; Kim, Soon Young

    2012-01-01

    The Core Protection Calculator System (CPCS) is an automated device which is adopted to inspect the safety parameters such as Departure from Nuclear Boiling Ratio (DNBR) and Local Power Density (LPD) during normal operation. One function of the CPCS is to predict the axial power distributions using function sets in cubic spline method. Another function of that is to impose penalty when the estimated distribution by the spline method disagrees with embedded data in CPCS (i.e., over 8%). In conventional CPCS, restricted function sets are used to synthesize axial power shape, whereby it occasionally can draw a disagreement between synthesized data and the embedded data. For this reason, the study on improvement for power distributions synthesis in CPCS has been conducted in many countries. In this study, many function sets (more than 18,000 types) differing from the conventional ones were evaluated in each power shape. Matlab code was used for calculating/arranging the numerous cases of function sets. Their synthesis performance was also evaluated through error between conventional data and consequences calculated by new function sets

  18. Effects of nuclear data library on BFS and ZPPR fast reactor core analysis results. Pt. 2. BFS-62 analysis results

    International Nuclear Information System (INIS)

    Mantourov, Guennadi

    2001-11-01

    This work was fulfilled in the frame of JNC-IPPE Collaboration on Experimental Investigation of Excess Weapon Pu Disposition in BN-600 Reactor Using BFS-2 Facility. Data processing system CONSYST/ABBN coupled with ABBN-93 nuclear data library was used in analysis of BFS-62 and ZPPR JUPIER series fast reactor cores, applying JNC core calculation code CITATION-FBR. FFCP cell code was used for taking into account the spatial cell heterogeneity and resonance effects based on the First Flight Collision Probability method and subgroup approach. Especially, two converting programs were written to transmit the prepared effective cross sections to JNC standard PDS files to let then the CITATION code be applied for 3-D HEXZ neutronics calculations of the investigated cores. The effects of nuclear data library have been studied by comparing the results calculated using ABBN-93 nuclear data library with the former ones obtained in JNC based on JENDL-3.2 nuclear data library. The comparison results using IPPE and JNC nuclear data libraries for k-effective parameter for 4 BFS-62 cores as well as for 3 ZPPR JUPITER experiment series cores ZPPR-9, ZPPR-13A and ZPPR-17A are presented. The comparison results for reaction rates distributions for 2 BFS-62 uranium loaded cores are included too. The calculated correction factors applied in all cases were less than 1.0%. The estimated uncertainty in k-effective C values caused by possible errors in calculation of the applied corrections is about 0.3% in case of BFS-62 and ZPPR MOX cores, and is about 0.2% for BFS-62 uranium-loaded cores. The main result of this study is that the effect of applying ABBN-93 nuclear data in JNC's calculation route for k-effective results is about 0.3% for ZPPR and BFS-62 cores with plutonium. As for BFS uranium-loaded cores (BFS-62-1 and BFS-62-2) the nuclear data library effect is about 0.1%. Next the sensitivity analysis was applied. It shown that the main contributors to the nuclear data library effect

  19. Development of Optimized Core Design and Analysis Methods for High Power Density BWRs

    Science.gov (United States)

    Shirvan, Koroush

    Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between ~he fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x16 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The plant systems outside the vessel are assumed to be the same as the ABWR-Il design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULA TE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater

  20. Optimization of extinguishing agents for nuclear power plants

    International Nuclear Information System (INIS)

    Boleman, M.; Lipar, M.; Balog, K.

    1998-01-01

    Focus is placed on use of extinguishing agents in nuclear power plants. The advantages and disadvantages of these agents are compared. Further perspectives for using particular extinguishing agents in nuclear power plants are outlined. (author)

  1. Preparation, process optimization and characterization of core-shell polyurethane/chitosan nanofibers as a potential platform for bioactive scaffolds.

    Science.gov (United States)

    Maleknia, Laleh; Dilamian, Mandana; Pilehrood, Mohammad Kazemi; Sadeghi-Aliabadi, Hojjat; Hekmati, Amir Houshang

    2018-06-01

    In this paper, polyurethane (PU), chitosan (Cs)/polyethylene oxide (PEO), and core-shell PU/Cs nanofibers were produced at the optimal processing conditions using electrospinning technique. Several methods including SEM, TEM, FTIR, XRD, DSC, TGA and image analysis were utilized to characterize these nanofibrous structures. SEM images exhibited that the core-shell PU/Cs nanofibers were spun without any structural imperfections at the optimized processing conditions. TEM image confirmed the PU/Cs core-shell nanofibers were formed apparently. It that seems the inclusion of Cs/PEO to the shell, did not induce the significant variations in the crystallinity in the core-shell nanofibers. DSC analysis showed that the inclusion of Cs/PEO led to the glass temperature of the composition increased significantly compared to those of neat PU nanofibers. The thermal degradation of core-shell PU/Cs was similar to PU nanofibers degradation due to the higher PU concentration compared to other components. It was hypothesized that the core-shell PU/Cs nanofibers can be used as a potential platform for the bioactive scaffolds in tissue engineering. Further biological tests should be conducted to evaluate this platform as a three dimensional scaffold with the capabilities of releasing the bioactive molecules in a sustained manner.

  2. Accumulating Data to Optimally Predict Obesity Treatment (ADOPT) Core Measures: Psychosocial Domain.

    Science.gov (United States)

    Sutin, Angelina R; Boutelle, Kerri; Czajkowski, Susan M; Epel, Elissa S; Green, Paige A; Hunter, Christine M; Rice, Elise L; Williams, David M; Young-Hyman, Deborah; Rothman, Alexander J

    2018-04-01

    Within the Accumulating Data to Optimally Predict obesity Treatment (ADOPT) Core Measures Project, the psychosocial domain addresses how psychosocial processes underlie the influence of obesity treatment strategies on weight loss and weight maintenance. The subgroup for the psychosocial domain identified an initial list of high-priority constructs and measures that ranged from relatively stable characteristics about the person (cognitive function, personality) to dynamic characteristics that may change over time (motivation, affect). This paper describes (a) how the psychosocial domain fits into the broader model of weight loss and weight maintenance as conceptualized by ADOPT; (b) the guiding principles used to select constructs and measures for recommendation; (c) the high-priority constructs recommended for inclusion; (d) domain-specific issues for advancing the science; and (e) recommendations for future research. The inclusion of similar measures across trials will help to better identify how psychosocial factors mediate and moderate the weight loss and weight maintenance process, facilitate research into dynamic interactions with factors in the other ADOPT domains, and ultimately improve the design and delivery of effective interventions. © 2018 The Obesity Society.

  3. Accumulating Data to Optimally Predict Obesity Treatment (ADOPT) Core Measures: Environmental Domain.

    Science.gov (United States)

    Saelens, Brian E; Arteaga, S Sonia; Berrigan, David; Ballard, Rachel M; Gorin, Amy A; Powell-Wiley, Tiffany M; Pratt, Charlotte; Reedy, Jill; Zenk, Shannon N

    2018-04-01

    There is growing interest in how environment is related to adults' weight and activity and eating behaviors. However, little is known about whether environmental factors are related to the individual variability seen in adults' intentional weight loss or maintenance outcomes. The environmental domain subgroup of the Accumulating Data to Optimally Predict obesity Treatment (ADOPT) Core Measures Project sought to identify a parsimonious set of objective and perceived neighborhood and social environment constructs and corresponding measures to include in the assessment of response to adult weight-loss treatment. Starting with the home address, the environmental domain subgroup recommended for inclusion in future weight-loss or maintenance studies constructs and measures related to walkability, perceived land use mix, food outlet accessibility (perceived and objective), perceived food availability, socioeconomics, and crime-related safety (perceived and objective) to characterize the home neighborhood environment. The subgroup also recommended constructs and measures related to social norms (perceived and objective) and perceived support to characterize an individual's social environment. The 12 neighborhood and social environment constructs and corresponding measures provide a succinct and comprehensive set to allow for more systematic examination of the impact of environment on adults' weight loss and maintenance. © 2018 The Obesity Society.

  4. Evaluating the Efficiency of a Multi-core Aware Multi-objective Optimization Tool for Calibrating the SWAT Model

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, X. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Izaurralde, R. C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zong, Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhao, K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Thomson, A. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2012-08-20

    The efficiency of calibrating physically-based complex hydrologic models is a major concern in the application of those models to understand and manage natural and human activities that affect watershed systems. In this study, we developed a multi-core aware multi-objective evolutionary optimization algorithm (MAMEOA) to improve the efficiency of calibrating a worldwide used watershed model (Soil and Water Assessment Tool (SWAT)). The test results show that MAMEOA can save about 1-9%, 26-51%, and 39-56% time consumed by calibrating SWAT as compared with sequential method by using dual-core, quad-core, and eight-core machines, respectively. Potential and limitations of MAMEOA for calibrating SWAT are discussed. MAMEOA is open source software.

  5. Optimizing the dynamic response of the H.B. Robinson nuclear plant using multiobjective particle swarm optimization

    International Nuclear Information System (INIS)

    Elsays, Mostafa A.; Naguib Aly, M.; Badawi, Alya A.

    2009-01-01

    In this paper, the Particle Swarm Optimization (PSO) algorithm is modified to deal with Multiobjective Optimization Problems (MOPs). A mathematical model for predicting the dynamic response of the H. B. Robinson nuclear power plant, which represents an Initial Value Problem (IVP) of Ordinary Differential Equations (ODEs), is solved using Runge-Kutta formula. The resulted data values are represented as a system of nonlinear algebraic equations by interpolation schemes for data fitting. This system of fitted nonlinear algebraic equations represents a nonlinear multiobjective optimization problem. A Multiobjective Particle Swarm Optimizer (MOPSO) which is based on the Pareto optimality concept is developed and applied to maximize the above mentioned problem. Results show that MOPSO efficiently cope with the problem and finds multiple Pareto optimal solutions. (orig.)

  6. Utilization of cross-section covariance data in FBR core nuclear design and cross-section adjustment

    International Nuclear Information System (INIS)

    Ishikawa, Makoto

    1994-01-01

    In the core design of large fast breeder reactors (FBRs), it is essentially important to improve the prediction accuracy of nuclear characteristics from the viewpoint of both reducing cost and insuring reliability of the plant. The cross-section errors, that is, covariance data are one of the most dominant sources for the prediction uncertainty of the core parameters, therefore, quantitative evaluation of covariance data is indispensable for FBR core design. The first objective of the present paper is to introduce how the cross-section covariance data are utilized in the FBR core nuclear design works. The second is to delineate the cross-section adjustment study and its application to an FBR design, because this improved design method markedly enhances the needs and importance of the cross-section covariance data. (author)

  7. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  8. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

  9. Fuel/propellant mixing in an open-cycle gas core nuclear rocket engine

    International Nuclear Information System (INIS)

    Guo, X.; Wehrmeyer, J.A.

    1997-01-01

    A numerical investigation of the mixing of gaseous uranium and hydrogen inside an open-cycle gas core nuclear rocket engine (spherical geometry) is presented. The gaseous uranium fuel is injected near the centerline of the spherical engine cavity at a constant mass flow rate, and the hydrogen propellant is injected around the periphery of the engine at a five degree angle to the wall, at a constant mass flow rate. The main objective is to seek ways to minimize the mixing of uranium and hydrogen by choosing a suitable injector geometry for the mixing of light and heavy gas streams. Three different uranium inlet areas are presented, and also three different turbulent models (k-var-epsilon model, RNG k-var-epsilon model, and RSM model) are investigated. The commercial CFD code, FLUENT, is used to model the flow field. Uranium mole fraction, axial mass flux, and radial mass flux contours are obtained. copyright 1997 American Institute of Physics

  10. FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores

    International Nuclear Information System (INIS)

    Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.

    1995-01-01

    FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs

  11. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    International Nuclear Information System (INIS)

    Downey, G.L.

    1988-01-01

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action. (author)

  12. How did Fukushima-Dai-ichi core meltdown change the probability of nuclear accidents?

    International Nuclear Information System (INIS)

    Escobar Rangel, Lina; Leveque, Francois

    2012-10-01

    How to predict the probability of a nuclear accident using past observations? What increase in probability the Fukushima Dai-ichi event does entail? Many models and approaches can be used to answer these questions. Poisson regression as well as Bayesian updating are good candidates. However, they fail to address these issues properly because the independence assumption in which they are based on is violated. We propose a Poisson Exponentially Weighted Moving Average (PEWMA) based in a state-space time series approach to overcome this critical drawback. We find an increase in the risk of a core meltdown accident for the next year in the world by a factor of ten owing to the new major accident that took place in Japan in 2011. (authors)

  13. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Downey, G L

    1988-05-01

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action.

  14. Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

    International Nuclear Information System (INIS)

    Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

    2008-01-01

    In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

  15. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    International Nuclear Information System (INIS)

    Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

  16. Hualong One's nuclear reactor core design and relative safety issues research

    Energy Technology Data Exchange (ETDEWEB)

    Yu, H., E-mail: yuhong_xing@126.com [Nuclear Power Inst. of China, Design and Research Sub-Inst., Chengdu, Sichuan (China)

    2015-07-01

    'Full text:' Hualong One, a third generation 1000MWe-class pressurized water reactor, is developed by China National Nuclear Cooperation (CNNC), based on the self-reliant technologies and experiences from China 40 years designing, construction, operation and maintenance of NPPs. In China, it has been approved to construct at Fuqing 5&6 and Fangchenggang 3&4. The Hualong One adopts advanced design features to dramatically enhance plant safety, economic efficiency and convenience of operation and maintenance. It consists of three loops with nominal thermal power output 3060 MWt and a 60-year design life. Its reactor core has 177 fuel assemblies, 18 month refueling interval (after initial cycle), and more than 15% thermal margin. It adopts low leakage loading pattern which can achieve better economy of the neutron, higher reactivity and lower radiation damage of pressure vessel. For the safety design, incorporating the feedback of Fukushima accident, the Hualong One has a combination of active and passive safety systems, a single station layout, double containment structure, and comprehensive implementation of defence-in-depth design principles. The new design features has been successfully evaluated to ensure that they enhance the performance and safety of Hualong One. Several experimental activates have been conducted, such as cavity injection and cooling system testing, passive containment heat removal system testing, and passive residual heat removal system of secondary side testing. The future improvements of Hualong reactor will focus on better economic core design and more reliable safety system. (author)

  17. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

  18. Assessment of water hammer effects on boiling water nuclear reactor core dynamics

    Directory of Open Access Journals (Sweden)

    Bousbia-Salah Anis

    2007-01-01

    Full Text Available Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .

  19. Development of advanced nuclear core analysis system applicable to various reactor types (II)

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2003-03-01

    A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

  20. Optimization of radially heterogeneous 1000-MW(e) LMFBR core configurations. Appendixes D and E. Research project 620-25

    International Nuclear Information System (INIS)

    Barthold, W.P.; Orechwa, Y.; Su, S.F.; Hutter, E.; Batch, R.V.; Beitel, J.C.; Turski, R.B.; Lam, P.S.K.

    1979-11-01

    A parameter study was conducted to determine the interrelated effects of: loosely or tightly coupled fuel regions separated by internal blanket assemblies, number of fuel regions, core height, number and arrangement of internal blanket subassemblies, number and size of fuel pins in a subassembly, etc. the effects of these parameters on sodium void reactivity, Doppler, incoherence, breeding gain, and thermohydraulics were of prime interest. Trends were established and ground work laid for optimization of a large, radially-heterogeneous, LMFBR core that will have low energetics in an HCDA and will have good thermal and breeding performance

  1. Optimizing the updated Goddard shortwave radiation Weather Research and Forecasting (WRF) scheme for Intel Many Integrated Core (MIC) architecture

    Science.gov (United States)

    Mielikainen, Jarno; Huang, Bormin; Huang, Allen H.-L.

    2015-05-01

    Intel Many Integrated Core (MIC) ushers in a new era of supercomputing speed, performance, and compatibility. It allows the developers to run code at trillions of calculations per second using the familiar programming model. In this paper, we present our results of optimizing the updated Goddard shortwave radiation Weather Research and Forecasting (WRF) scheme on Intel Many Integrated Core Architecture (MIC) hardware. The Intel Xeon Phi coprocessor is the first product based on Intel MIC architecture, and it consists of up to 61 cores connected by a high performance on-die bidirectional interconnect. The co-processor supports all important Intel development tools. Thus, the development environment is familiar one to a vast number of CPU developers. Although, getting a maximum performance out of Xeon Phi will require using some novel optimization techniques. Those optimization techniques are discusses in this paper. The results show that the optimizations improved performance of the original code on Xeon Phi 7120P by a factor of 1.3x.

  2. Quantitative Proteomic Analysis of Optimal Cutting Temperature (OCT) Embedded Core-Needle Biopsy of Lung Cancer

    Science.gov (United States)

    Zhao, Xiaozheng; Huffman, Kenneth E.; Fujimoto, Junya; Canales, Jamie Rodriguez; Girard, Luc; Nie, Guangjun; Heymach, John V.; Wistuba, Igacio I.; Minna, John D.; Yu, Yonghao

    2017-10-01

    With recent advances in understanding the genomic underpinnings and oncogenic drivers of pathogenesis in different subtypes, it is increasingly clear that proper pretreatment diagnostics are essential for the choice of appropriate treatment options for non-small cell lung cancer (NSCLC). Tumor tissue preservation in optimal cutting temperature (OCT) compound is commonly used in the surgical suite. However, proteins recovered from OCT-embedded specimens pose a challenge for LC-MS/MS experiments, due to the large amounts of polymers present in OCT. Here we present a simple workflow for whole proteome analysis of OCT-embedded NSCLC tissue samples, which involves a simple trichloroacetic acid precipitation step. Comparisons of protein recovery between frozen versus OCT-embedded tissue showed excellent consistency with more than 9200 proteins identified. Using an isobaric labeling strategy, we quantified more than 5400 proteins in tumor versus normal OCT-embedded core needle biopsy samples. Gene ontology analysis indicated that a number of proliferative as well as squamous cell carcinoma (SqCC) marker proteins were overexpressed in the tumor, consistent with the patient's pathology based diagnosis of "poorly differentiated SqCC". Among the most downregulated proteins in the tumor sample, we noted a number of proteins with potential immunomodulatory functions. Finally, interrogation of the aberrantly expressed proteins using a candidate approach and cross-referencing with publicly available databases led to the identification of potential druggable targets in DNA replication and DNA damage repair pathways. We conclude that our approach allows LC-MS/MS proteomic analyses on OCT-embedded lung cancer specimens, opening the way to bring powerful proteomics into the clinic. [Figure not available: see fulltext.

  3. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  4. An optimization method for parameters in reactor nuclear physics

    International Nuclear Information System (INIS)

    Jachic, J.

    1982-01-01

    An optimization method for two basic problems of Reactor Physics was developed. The first is the optimization of a plutonium critical mass and the bruding ratio for fast reactors in function of the radial enrichment distribution of the fuel used as control parameter. The second is the maximization of the generation and the plutonium burnup by an optimization of power temporal distribution. (E.G.) [pt

  5. Development of intelligent code system to support conceptual design of nuclear reactor core

    International Nuclear Information System (INIS)

    Kugo, Teruhiko; Nakagawa, Masayuki; Tsuchihashi, Keichiro

    1997-01-01

    An intelligent reactor design system IRDS has been developed to support conceptual design of new type reactor cores in the fields of neutronics, thermal-hydraulics and fuel behavior. The features of IRDS are summarized as follows: 1) a variety of computer codes to cover various design tasks relevant to 'static' and 'burnup' problems are implemented, 2) all the information necessary to the codes implemented is unified in a data base, 3) several data and knowledge bases are referred to in order to proceed design process efficiently for non-expert users, 4) advanced man-machine interface to communicate with the system through an interactive and graphical user interface is equipped and 5) a function to search automatically a design window, which is defined as a feasible parameter range to satisfy design requirement and criteria is employed to support the optimization or satisfication process. Applicability and productivity of the system are demonstrated by the design study of fuel pin for new type FBR cores. (author)

  6. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  7. Optimal usage of computing grid network in the fields of nuclear fusion computing task

    International Nuclear Information System (INIS)

    Tenev, D.

    2006-01-01

    Nowadays the nuclear power becomes the main source of energy. To make its usage more efficient, the scientists created complicated simulation models, which require powerful computers. The grid computing is the answer to powerful and accessible computing resources. The article observes, and estimates the optimal configuration of the grid environment in the fields of the complicated nuclear fusion computing tasks. (author)

  8. Optimizing Sparse Matrix-Multiple Vectors Multiplication for Nuclear Configuration Interaction Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Aktulga, Hasan Metin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Buluc, Aydin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Williams, Samuel [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Yang, Chao [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2014-08-14

    Obtaining highly accurate predictions on the properties of light atomic nuclei using the configuration interaction (CI) approach requires computing a few extremal Eigen pairs of the many-body nuclear Hamiltonian matrix. In the Many-body Fermion Dynamics for nuclei (MFDn) code, a block Eigen solver is used for this purpose. Due to the large size of the sparse matrices involved, a significant fraction of the time spent on the Eigen value computations is associated with the multiplication of a sparse matrix (and the transpose of that matrix) with multiple vectors (SpMM and SpMM-T). Existing implementations of SpMM and SpMM-T significantly underperform expectations. Thus, in this paper, we present and analyze optimized implementations of SpMM and SpMM-T. We base our implementation on the compressed sparse blocks (CSB) matrix format and target systems with multi-core architectures. We develop a performance model that allows us to understand and estimate the performance characteristics of our SpMM kernel implementations, and demonstrate the efficiency of our implementation on a series of real-world matrices extracted from MFDn. In particular, we obtain 3-4 speedup on the requisite operations over good implementations based on the commonly used compressed sparse row (CSR) matrix format. The improvements in the SpMM kernel suggest we may attain roughly a 40% speed up in the overall execution time of the block Eigen solver used in MFDn.

  9. PSA-based optimization of technical specifications for the Borssele nuclear power plant

    International Nuclear Information System (INIS)

    Seebregts, A.J.; Schoonakker, H.A.

    1996-01-01

    The Borssele Nuclear Power Plant (NPP) is a Siemens/KWU 472 MWe Pressurized Water Reactor which has been in operation since 1973. In 1989, a Probabilistic Safety Assessment (PSA) program was initiated to complement deterministic safety studies and operational experience in forming a plant safety concept. In 1993, the PSA-MER model was completed and used to determine the effects a package of proposed modifications would have on plant safety and risks to the environment. This model was used to start retrospective risks profile and allowed outage times (AOTs) analyses, which both concerned the calculation of the change in total core damage frequency (TCDF) given a change in configuration. The main problems identified and reported in this paper are: (i) How to calculate the change in TCDF (ΔTCDF)? (section 3); and (ii) How to set practical decision criteria and how to use the PSA as extension to Technical Specifications (TS) AOTs? (section 4). Finally, a pilot study was conducted in order to optimize surveillance test intervals (STIs) which are also part of the TS (section 5). (orig.)

  10. Contribution to the evaluation and to the improvement of multi-objective optimization methods: application to the optimization of nuclear fuel reloading pattern

    International Nuclear Information System (INIS)

    Collette, Y.

    2002-01-01

    In this thesis, we study the general problem of the selection of a multi-objective optimization method, then we study the improvement so as to efficiently solve a problem. The pertinent selection of a method presume the existence of a methodology: we have built tools to perform evaluation of performances and we propose an original method dedicated to the classification of know optimization methods. Our step has been applied to the elaboration of new methods for solving a very difficult problem: the nuclear core reload pattern optimization. First, we looked for a non usual approach of performances measurement: we have 'measured' the behavior of a method. To reach this goal, we have introduced several metrics. We have proposed to evaluate the 'aesthetic' of a distribution of solutions by defining two new metrics: a 'spacing metric' and a metric that allow us to measure the size of the biggest hole in the distribution of solutions. Then, we studied the convergence of multi-objective optimization methods by using some metrics defined in scientific literature and by proposing some more metrics: the 'Pareto ratio' which computes a ratio of solution production. Lastly, we have defined new metrics intended to better apprehend the behavior of optimization methods: the 'speed metric', which allows to compute the speed profile and a 'distribution metric' which allows to compute statistical distribution of solutions along the Pareto frontier. Next, we have studied transformations of a multi-objective problem and defined news methods: the modified Tchebychev method, or the penalized weighted sum of objective functions. We have elaborated new techniques to choose the initial point. These techniques allow to produce new initial points closer and closer to the Pareto frontier and, thanks to the 'proximal optimality concept', allowing dramatic improvements in the convergence of a multi-objective optimization method. Lastly, we have defined new vectorial multi-objective optimization

  11. Integrated approach to optimize operation and maintenance costs for operating nuclear power plants

    International Nuclear Information System (INIS)

    2006-06-01

    In the context of increasingly open electricity markets and the 'unbundling' of generating companies from former utility monopolies, an area of major concern is the economic performance of the existing fleet of nuclear power plants. Nuclear power, inevitably, must compete directly with other electricity generation sources. Coping with this competitive pressure is a challenge that the nuclear industry should meet if the nuclear option is to remain a viable one. This competitive environment has significant implications for nuclear plant operations, including, among others, the need for the more cost effective management of plant activities, and the greater use of analytical tools to balance the costs and benefits of proposed activities, in order to optimize operation and maintenance costs, and thus insure the economic competitiveness of existing nuclear power plants. In the framework of the activities on Nuclear Economic Performance Information System (NEPIS), the IAEA embarked in developing guidance on optimization of operation and maintenance costs for nuclear power plants. The report was prepared building on the fundamental that optimization of operation and maintenance costs of a nuclear power plant is a key component of a broader integrated business strategic planning process, having as overall result achievement of organization's business objectives. It provides advice on optimization of O and M costs in the framework of strategic business planning, with additional details on operational planning and controlling. This TECDOC was elaborated in 2004-2005 in the framework of the IAEA's programme on Nuclear Power Plant Operating Performance and Life Cycle Management, with the support of two consultants meetings and one technical meeting and based on contributions provided by participants. It can serve as a useful reference for the management and operation staff within utilities, nuclear power plant operators and regulators and other organizations involved in

  12. Optimization and Openmp Parallelization of a Discrete Element Code for Convex Polyhedra on Multi-Core Machines

    Science.gov (United States)

    Chen, Jian; Matuttis, Hans-Georg

    2013-02-01

    We report our experiences with the optimization and parallelization of a discrete element code for convex polyhedra on multi-core machines and introduce a novel variant of the sort-and-sweep neighborhood algorithm. While in theory the whole code in itself parallelizes ideally, in practice the results on different architectures with different compilers and performance measurement tools depend very much on the particle number and optimization of the code. After difficulties with the interpretation of the data for speedup and efficiency are overcome, respectable parallelization speedups could be obtained.

  13. Towards a Systematic Exploration of the Optimization Space for Many-Core Processors

    NARCIS (Netherlands)

    Fang, J.

    2014-01-01

    The architecture diversity of many-core processors - with their different types of cores, and memory hierarchies - makes the old model of reprogramming every application for every platform infeasible. Therefore, inter-platform portability has become a desirable feature of programming models. While

  14. ROSA full-core and DNBR capabilities

    International Nuclear Information System (INIS)

    Gibcus, H.P.M.; Verhagen, F.C.M.; Wakker, P.H.

    2013-01-01

    The latest developments of the ROSA (Reloading Optimization by Simulated Annealing) code system with an emphasis on the first full-core version and the minimum DNBR (Departure from Nucleate Boiling Ratio) as a new optimization parameter are presented. Designing the core loading pattern of nuclear power plants is becoming a more and more complex task. This task becomes even more complicated if asymmetries in the core loading pattern arise, for instance due to damaged fuel assemblies. For over almost 2 decades ROSA, NRG's (Nuclear Research and consultancy Group) loading pattern optimization code system for PWRs, has proven to be a valuable tool to reactor operators in accomplishing this task. To improve the use of ROSA for designing asymmetric loading patterns, NRG has developed a full-core version of ROSA besides the original quarter-core version which requires rotational symmetry in the computational domain. The extension of ROSA with DNBR as an optimization parameter is part of ROSA's continuous development. (orig.)

  15. Further optimization of the M1 PAM VU0453595: Discovery of novel heterobicyclic core motifs with improved CNS penetration.

    Science.gov (United States)

    Panarese, Joseph D; Cho, Hykeyung P; Adams, Jeffrey J; Nance, Kellie D; Garcia-Barrantes, Pedro M; Chang, Sichen; Morrison, Ryan D; Blobaum, Anna L; Niswender, Colleen M; Stauffer, Shaun R; Conn, P Jeffrey; Lindsley, Craig W

    2016-08-01

    This Letter describes the continued chemical optimization of the VU0453595 series of M1 positive allosteric modulators (PAMs). By surveying alternative 5,6- and 6,6-heterobicylic cores for the 6,7-dihydro-5H-pyrrolo[3,4-b]pyridine-5-one core of VU453595, we found new cores that engendered not only comparable or improved M1 PAM potency, but significantly improved CNS distribution (Kps 0.3-3.1). Moreover, this campaign provided fundamentally distinct M1 PAM chemotypes, greatly expanding the available structural diversity for this valuable CNS target, devoid of hydrogen-bond donors. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. Optimization for set-points and robust model predictive control for steam generator in nuclear power plants

    International Nuclear Information System (INIS)

    Osgouee, Ahmad

    2010-01-01

    Full Text: Nuclear power plants will be needed for future energy demands, which are expected to grow at different rates around the world. Lower operating cost is one of the major benefits of nuclear power plants over fossil power plants. Also, the plant availability is a key factor to economic index of a nuclear power plant. The opportunities for building new nuclear power plants around the world will depend on the need for clean energy with zero, or minimal emissions to support healthy communities, supply reliable energy with stable prices, and issues related to global warming and climate change. Compared to other types of power plants, nuclear power plants are preferred for their numerous advantages, including low operating costs, emission free operation with no smog, no acid rain, and no effect on global warming. Economic feasibility of a nuclear power plant requires for smooth and uninterrupted plant operation during electrical power demand variations. The steam generator (SG) in a nuclear power plant plays an important role in cooling of the reactor, balancing energy between reactor and turbine and producing steam for the turbine-generators. SG acts as an additional safety barrier between the nuclear reactor and the outside world also. As a result, control of the water inventory in the SG is very important to ensure continuous cooling of the nuclear reactor core, plant protection and at the same time, to prevent the SG tubes and turbine blades failure. A review of past nuclear power plant operation experiences indicates that unplanned reactor trips due to steam generator level (SGL) control have been significant contributors to plant unavailability. During low power operation, the level control is complicated by the thermal reverse effects known as 'shrink and swell'. Manual operator intervention to the SGL control system at low reactor power and to the unit upset conditions has been identified as an operator response in most nuclear power plants. In spite of

  17. Study on a nuclear spaceship for interplanetary cruise. Core design of a small fast reactor

    International Nuclear Information System (INIS)

    Kitamura, Taku; Yoshida, Yutaka; Honma, Yuji; Narabayashi, Tadashi; Shimazu, Yoichiro; Tsuji, Masashi

    2009-01-01

    In 21st century, the field which needs nuclear power plant systems are not just on the Earth. We considered that the nuclear power is proper for the energy source of the manned spaceship for interplanetary cruise. In this study, we considered the system configuration of the spaceship, the design of power generating system, some navigational plans to reach the Mars. The system configuration of the spaceship studied in our laboratory has one or two Fast Reactor with liquid sodium coolant as main heat source, dozens of Stirling Engines as main power generators and some Plasma Rockets called VASIMR as propulsion system. Because the Fast Reactor need not thick and heavy pressure vessel and the sodium has high performance of heat transfer, they are the best suited to the space nuclear reactor system. In addition, Stirling Engine has high theoretical thermal efficiency and need not water, steam generators, steam condenser and so on. This results in absence of sodium-water reaction and significant weight saving of power generator system. The VASIMR studied at ASPL is an advanced electric propulsion device which is able to convert large amount of electric power into great propulsion force. At reactor designing, we are using the SRAC2006 code developed at JAEA and pursuing the optimal fast reactor design for spaceship. We think that smaller reactor is better. To realize a system which has inherent safety, sodium void reactivity should be negative. We adopted the design of the small fast reactor named 4S (Super Safe, Small and Simple) as a reference design. As a result, we verified that a void reactivity had negative value in some of calculation cases and we realized safe, small and simple space fast reactor. In addition, to piece out power generator system in space, we need to consider if the budget of exhaust heat from radiator panels to space needed at this case is realistic. To obtain the optimal trajectory of rapid Mars transit, we made some analysis calculation codes

  18. Sample size optimization in nuclear material control. 1

    International Nuclear Information System (INIS)

    Gladitz, J.

    1982-01-01

    Equations have been derived and exemplified which allow the determination of the minimum variables sample size for given false alarm and detection probabilities of nuclear material losses and diversions, respectively. (author)

  19. Optimization and parallelization of B-spline based orbital evaluations in QMC on multi/many-core shared memory processors

    OpenAIRE

    Mathuriya, Amrita; Luo, Ye; Benali, Anouar; Shulenburger, Luke; Kim, Jeongnim

    2016-01-01

    B-spline based orbital representations are widely used in Quantum Monte Carlo (QMC) simulations of solids, historically taking as much as 50% of the total run time. Random accesses to a large four-dimensional array make it challenging to efficiently utilize caches and wide vector units of modern CPUs. We present node-level optimizations of B-spline evaluations on multi/many-core shared memory processors. To increase SIMD efficiency and bandwidth utilization, we first apply data layout transfo...

  20. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.