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Sample records for fuel reloading sequence

  1. 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)

  2. An optimal sequence of the reload charge fuel enrichment to a reactor

    International Nuclear Information System (INIS)

    Sato, S.

    1975-01-01

    An optimal sequence of enrichment of the reload charge of a three regions PWR during its life has been determined by dynamic programming. The state of the reactor is specified by the burnup of the fuel in the three regions and their initial enrichments. Constraints were imposed on the power peaking factor, the maximum burnup, the length of each stage between refueling and the total life of the reactor. 'Central-scatter loading' was assumed at each reloading. The two group diffusion equations were solved by the modal method for the static calculations of the reactor. Otimization of enrichment of the reload charge was performed under several hypotheses on the variation of the costs of uranium, costs of enrichment and the plant factor during the reactor life. It was observed that the optimum enrichment of the reload fuel is influenced more by the cost of enrichment rather than plant factor or cost of uranium. (Author) [pt

  3. Nuclear design of APSARA reload-2 fuel

    International Nuclear Information System (INIS)

    Nath, M.; Veeraraghavan, N.

    1978-01-01

    In view of the satisfactory operating performance of initial and reload-1 fuel designs of Apsara reactor, it was felt desirable to adopt a basically similar design for reload-2 fuel, i.e. the fuel assembly should consist of equally spaced parallel fuel plates in which highly enriched uranium, alloyed with aluminium, is employed as fuel. However, because of fabricational constraints, certain modifications were necessary and were incorporated in the proposed reload design to cater to the multiple needs of operational requirements, improved fuel utilization and inherent reactor safety. The salient features of the nuclear design of reload-2 fuel for the Apsara reactor are discussed. (author)

  4. TAPS safety evaluation criteria for reload fueling

    International Nuclear Information System (INIS)

    Mahendra Nath; Veeraraghavan, N.

    1976-01-01

    To improve operating performance of Tarapur reactors, several proposals are under consideration such as core expansion, change-over to an improved fuel design with lower heat rating, extension of fuel cycle lengths etc., which have a bearing on overall plant operating characteristics and reactor safety. For evaluating safety implications of the various proposals, it is necessary to formulate safety evaluation criteria for reload fuelling. Salient features of these criteria are discussed. (author)

  5. First fuel reload in Laguna Verde

    International Nuclear Information System (INIS)

    Bahena B, D.

    1992-01-01

    A report containing the activities carried out during the first reload of nuclear fuel and major maintenance in the Laguna Verde nuclear reactor is presented. The previous and the specific activities are included. These last are related without including the technical considerations, data or the operation details, because these data were documented inside the registrations of the CFE, the ININ and in personal way. (Author)

  6. ANTQ evolutionary algorithm applied to nuclear fuel reload problem

    International Nuclear Information System (INIS)

    Machado, Liana; Schirru, Roberto

    2000-01-01

    Nuclear fuel reload optimization is a NP-complete combinatorial optimization problem where the aim is to find fuel rods' configuration that maximizes burnup or minimizes the power peak factor. For decades this problem was solved exclusively using an expert's knowledge. From the eighties, however, there have been efforts to automatize fuel reload. The first relevant effort used Simulated Annealing, but more recent publications show Genetic Algorithm's (GA) efficiency on this problem's solution. Following this direction, our aim is to optimize nuclear fuel reload using Ant-Q, a reinforcement learning algorithm based on the Cellular Computing paradigm. Ant-Q's results on the Travelling Salesmen Problem, which is conceptually similar to fuel reload, are better than the GA's ones. Ant-Q was tested on fuel reload by the simulation of the first cycle in-out reload of Bibils, a 193 fuel element PWR. Comparing An-Q's result with the GA's ones, it can b seen that even without a local heuristics, the former evolutionary algorithm can be used to solve the nuclear fuel reload problem. (author)

  7. A study for fuel reloading strategy in pebble bed core

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2012-02-01

    A fuel reloading analysis system for pebble bed reactor was developed by using a Monte Carlo code. The kinematic model was modified to improve the accuracy of the pebble velocity profile and to develop the model so that the diffusion coefficient is not changed by the geometry of the core. In addition, the point kernel method was employed to solve an equation derived in this study. Then, the analysis system for the pebble bed reactor was developed to accommodate the double heterogeneity, pebble velocity, and pebble refueling features using the MCNPX Monte Carlo code. The batch-tracking method was employed to simulate the movement of the pebbles and an automation system was written in the C programming language to implement it. The proposed analysis system can be utilized to verify new core analysis codes, deep-burn studies, various sensitivity studies, and other analysis tools available for the application of new fuel reloading strategies. It is noted that the proposed algorithm for the optimum fuel reloading pattern differs from other optimization methods using sensitivity analysis. In this algorithm, the reloading strategy, including the loading of fresh fuel and the reloading positions of the fresh and reloaded fuels, is determined by the interrelations of the criticality, the nuclear material inventories in the extracted fuel, and the power density. The devised algorithm was applied to the PBMR and NHDD-PBR200. The results show that the proposed algorithm can apply to satisfy the nuclear characteristics such as the criticality or power density since the pebble bed core has the characteristics that the fuels are reloaded every day

  8. Economic study of fuel scenarios for a reload

    International Nuclear Information System (INIS)

    Ortiz S, J. J.; Castillo M, J. A.; Montes T, J. L.; Perusquia del C, R.

    2014-10-01

    In this work the results to plan different scenarios for designing a nuclear fuel reload are shown. Given a reload with specific energy requirements, the objective is to verify the feasibility of using either a greater number of fresh fuel with less uranium enrichment, or otherwise reduce the number of fresh fuel assemblies and therefore they have a higher average uranium enrichment. For the study a cycle balance 18-month basis with 112 fresh assemblies divided into two lots, with energy produced of 10,075 Mwd/Tu was used. For the designs under the mentioned scenarios, the heuristic techniques known as taboo search and neural networks were used. To verify the feasibility of obtained reloads an economic study of the reload costs was performed. The results showed that is possible to design reloads under the two scenarios, but was more complicated decrease the amount of fresh fuel assemblies. In both scenarios was possible to reduce manufacturing costs of fuel and according to purely static calculation, it would be possible to increase the energy produced. (Author)

  9. From FUELCON to FUELGEN: tools for fuel reload pattern design

    International Nuclear Information System (INIS)

    Nissan, E.; Zhao, J.; Knight, B.; Soper, A.; Galperin, A.

    2000-01-01

    FUELGEN is an effective tool for refuelling design, i.e., for solving the incore fuel management problem at nuclear power plants. Devising good fuel allocations for reloading the core of a given nuclear reactor, for a given operation cycle, is crucial for keeping down operation costs at plants. Fuel comes in different types, and is positioned in a grid representing the core of a reactor. The starting point was Galperin and Nissan's prototype which eventually led to FUELCON, a rule-based expert system with the same task. FUELGEN, instead, is based on a genetic algorithm for optimization, and is at the current forefront of research in refuelling design, where genetic techniques are now getting increasing recognition. The end result of over a decade of research within this sequence of projects yielded a set of alternative, partly overlapping architectures. Nodal algorithms to carry out parameter prediction by simulation, heuristic rules in FUELCON's ruleset and metal-level refinement ergonomic techniques by which the ruleset can be refined during a session with FUELCON, attempts with neural computation on top of the latter, and then, replacing the ruleset altogether by resorting to genetic algorithms, are the sequence of techniques that were in turn applied, in the development of FUELCON and the FUELGEN. This actually reflects the sequence of emergence of expert systems and then neural computation methods, then genetic and hybrid methods, in knowledge engineering in general and in its application to nuclear engineering in particular. (orig.)

  10. First fuel re-load of Angra-1 reactor - Inspection and hearing plan

    International Nuclear Information System (INIS)

    Pollis, W.; Alvarenga, M.A.B.; Meldonian, N.L.; Paiva, R.L.C. de; Pollis, R.

    1985-01-01

    The plan of inspection and hearing of the first fuel reload of Angra-1 nuclear reactor is detailed. It consists in five steps: receiving and storage of the fuel; reload preparation; activities during; post-reload activities, and preliminary activities. (M.I.)

  11. Mixed Reload Design Using MOX and UOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Ramon, Ramirez Sanchez J.; Perry, R.T.

    2002-01-01

    As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10 X 10 BWR fuel assemblies but different fissile material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO 2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fissile plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO 2 fresh fuel were tested to verify the shutdown margin, the UO 2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper. (authors)

  12. 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)

  13. Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lan, Nguyen Phuoc; Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Binh, Do Quang [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs.

  14. 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)

  15. Design and optimization of a fuel reload of BWR with plutonium and minor actinides

    International Nuclear Information System (INIS)

    Guzman A, J. R.; Francois L, J. L.; Martin del Campo M, C.; Palomera P, M. A.

    2008-01-01

    In this work is designed and optimized a pattern of fuel reload of a boiling water reactor (BWR), whose fuel is compound of uranium coming from the enrichment lines, plutonium and minor actinides (neptunium, americium, curium); obtained of the spent fuel recycling of reactors type BWR. This work is divided in two stages: in the first stage a reload pattern designs with and equilibrium cycle is reached, where the reload lot is invariant cycle to cycle. This reload pattern is gotten adjusting the plutonium content of the assembly for to reach the length of the wished cycle. Furthermore, it is necessary to increase the concentration of boron-10 in the control rods and to introduce gadolinium in some fuel rods of the assembly, in order to satisfy the margin approach of out. Some reactor parameters are presented: the axial profile of power average of the reactor core, and the axial and radial distribution of the fraction of holes, for the one reload pattern in balance. For the design of reload pattern codes HELIOS and CM-PRESTO are used. In the second stage an optimization technique based on genetic algorithms is used, along with certain obtained heuristic rules of the engineer experience, with the intention of optimizing the reload pattern obtained in the first stage. The objective function looks for to maximize the length of the reactor cycle, at the same time as that they are satisfied their limits related to the power and the reactor reactivity. Certain heuristic rules are applied in order to satisfy the recommendations of the fuel management: the strategy of the control cells core, the strategy of reload pattern of low leakage, and the symmetry of a quarter of nucleus. For the evaluation of the parameters that take part in the objective function it simulates the reactor using code CM-PRESTO. Using the technique of optimization of the genetic algorithms an energy of the cycle of 10834.5 MW d/tHM is obtained, which represents 5.5% of extra energy with respect to the

  16. 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

  17. 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.

  18. Application of a heuristic search method for generation of fuel reload configurations

    International Nuclear Information System (INIS)

    Galperin, A.; Nissan, E.

    1988-01-01

    A computerized heuristic search method for the generation and optimization of fuel reload configurations is proposed and investigated. The heuristic knowledge is expressed modularly in the form of ''IF-THEN'' production rules. The method was implemented in a program coded in the Franz LISP programming language and executed under the UNIX operating system. A test problem was formulated, based on a typical light water reactor reload problem with a few simplifications assumed, in order to allow formulation of the reload strategy into a relatively small number of rules. A computer run of the problem was performed with a VAX-780 machine. A set of 312 solutions was generated in -- 20 min of execution time. Testing of a few arbitrarily chosen configurations demonstrated reasonably good performance for the computer-generated solutions. A computerized generator of reload configurations may be used for the fast generation or modification of reload patterns and as a tool for the formulation, tuning, and testing of the heuristic knowledge rules used by an ''expert'' fuel manager

  19. Design optimization for fuel reloading in Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    Cortes Campos, C.C.; Montes Tadeo, J.L.

    1991-01-01

    Procedure followed to perform the design optimation in fuel reloading is described in general words and also is shown an example in which such procedure was uses for analysis of BWR type reactor in unit 1 of Laguna Verde nuclear power plant (Author)

  20. Fuel assemblies with inert matrices as reloads of cycle 11 of the Unit 1 of the LVNC

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Hernandez M, N.; Hernandez L, H.

    2005-01-01

    In this work the results that were obtained of the analysis of three different reloads of the cycle 11 with fuel assemblies containing a mixture of UO 2 and plutonium grade armament in an inert matrix. The proposed assemble, consists of an arrangement 10x10 with 42 bars fuels of PuO 2 -CeO 2 , 34 fuel bars with UO 2 and 16 fuel bars with UO 2 -Gd 2O 3. The proposed assemble is equivalent to an it reloadable assemble of the cycle 11. The fuel bars of uranium and gadolinium, are of the same type of those that are used in the reloadable assemble of uranium. The design and generation of the nuclear databases of the fuel cell with mixed fuel, it was carried out with the HELIUMS code. The simulation of operation of the cycle 11, it was carried out with the CM-PRESTO code. The results show that with one reload of 72 assemblies of UO 2 and 32 assemblies with mixed fuel has a cycle length of smaller in 10.5 days to the cycle length with the complete reload of assemblies of UO 2 and a length smaller cycle in 34 days with the complete reload of 104 assemblies with mixed fuel. (Author)

  1. Optimization of fuel reloads for a BWR using the ant colony system

    International Nuclear Information System (INIS)

    Esquivel E, J.; Ortiz S, J. J.

    2009-10-01

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  2. An Order Coding Genetic Algorithm to Optimize Fuel Reloads in a Nuclear Boiling Water Reactor

    International Nuclear Information System (INIS)

    Ortiz, Juan Jose; Requena, Ignacio

    2004-01-01

    A genetic algorithm is used to optimize the nuclear fuel reload for a boiling water reactor, and an order coding is proposed for the chromosomes and appropriate crossover and mutation operators. The fitness function was designed so that the genetic algorithm creates fuel reloads that, on one hand, satisfy the constrictions for the radial power peaking factor, the minimum critical power ratio, and the maximum linear heat generation rate while optimizing the effective multiplication factor at the beginning and end of the cycle. To find the values of these variables, a neural network trained with the behavior of a reactor simulator was used to predict them. The computation time is therefore greatly decreased in the search process. We validated this method with data from five cycles of the Laguna Verde Nuclear Power Plant in Mexico

  3. First fuel reload in Laguna Verde; Primera recarga de combustible en Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Bahena B, D

    1992-01-15

    A report containing the activities carried out during the first reload of nuclear fuel and major maintenance in the Laguna Verde nuclear reactor is presented. The previous and the specific activities are included. These last are related without including the technical considerations, data or the operation details, because these data were documented inside the registrations of the CFE, the ININ and in personal way. (Author)

  4. A quality assurance programme for reload fuel for light-water reactors

    International Nuclear Information System (INIS)

    Nilson, R.

    1976-01-01

    The Exxon Nuclear quality assurance programme for the design and fabrication of reload fuel for light-water reactors is described. The programme is based on the 18 quality assurance criteria used for the design and construction of nuclear facilities in the United States of America, but is broadened considerably to reflect other inputs and experiences unique to nuclear fuel production. The government and utility interfaces with the fuel supplier in the area of quality assurance, and future trends, for example, the development of topical quality assurance reports, are also discussed. Quality assurance is discussed in terms of three fundamental categories: management control, engineering assurance and quality control. Examples of specific design, processing and inspection considerations which relate to known fuel failure mechanisms are discussed. The results of irradiated fuel examinations to date have shown that certain fuel failure mechanisms can be alleviated by the considerations described and that fuel of the requisite quality can be consistently produced. (author)

  5. Development and application of methods and computer codes of fuel management and nuclear design of reload cycles in PWR

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.; Corella, M.R.; Esteban, A.; Martinez-Val, J.M.; Minguez, E.; Perlado, J.M.; Pena, J.; Matias, E. de; Llorente, A.; Navascues, J.; Serrano, J.

    1976-01-01

    Description of methods and computer codes for Fuel Management and Nuclear Design of Reload Cycles in PWR, developed at JEN by adaptation of previous codes (LEOPARD, NUTRIX, CITATION, FUELCOST) and implementation of original codes (TEMP, SOTHIS, CICLON, NUDO, MELON, ROLLO, LIBRA, PENELOPE) and their application to the project of Management and Design of Reload Cycles of a 510 Mwt PWR, including comparison with results of experimental operation and other calculations for validation of methods. (author) [es

  6. Development of a graphical interface computer code for reactor fuel reloading optimization

    International Nuclear Information System (INIS)

    Do Quang Binh; Nguyen Phuoc Lan; Bui Xuan Huy

    2007-01-01

    This report represents the results of the project performed in 2007. The aim of this project is to develop a graphical interface computer code that allows refueling engineers to design fuel reloading patterns for research reactor using simulated graphical model of reactor core. Besides, this code can perform refueling optimization calculations based on genetic algorithms as well as simulated annealing. The computer code was verified based on a sample problem, which relies on operational and experimental data of Dalat research reactor. This code can play a significant role in in-core fuel management practice at nuclear research reactor centers and in training. (author)

  7. Selection of LWR cycle length and fuel reload fraction

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Handwerk, C.S.; McMahon, M.V.

    1997-01-01

    The continuing evolution of fuel having ever higher burnup capability and the increased emphasis on high plant capacity factor to keep nuclear power cost-competitive, motivates re-examination of some basic fuel management strategies. Specifically, what are the economic optimum goals for the fraction of core to be refueled, 1/n, and the length of the intra-refueling cycle, T c . The authors present a simple model to study these questions. They conclude that unless substantial improvements in technology are forthcoming, or economic circumstances change significantly, departure from 2- to 4-batch management, or longer than 2- to 3-year cycles in LWRs is not supported by their analysis

  8. 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

  9. Economic study of fuel scenarios for a reload; Estudio economico de escenarios de combustible para una recarga

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, J. A.; Montes T, J. L.; Perusquia del C, R., E-mail: juanjose.ortiz@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this work the results to plan different scenarios for designing a nuclear fuel reload are shown. Given a reload with specific energy requirements, the objective is to verify the feasibility of using either a greater number of fresh fuel with less uranium enrichment, or otherwise reduce the number of fresh fuel assemblies and therefore they have a higher average uranium enrichment. For the study a cycle balance 18-month basis with 112 fresh assemblies divided into two lots, with energy produced of 10,075 Mwd/Tu was used. For the designs under the mentioned scenarios, the heuristic techniques known as taboo search and neural networks were used. To verify the feasibility of obtained reloads an economic study of the reload costs was performed. The results showed that is possible to design reloads under the two scenarios, but was more complicated decrease the amount of fresh fuel assemblies. In both scenarios was possible to reduce manufacturing costs of fuel and according to purely static calculation, it would be possible to increase the energy produced. (Author)

  10. 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

  11. 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

  12. A knowledge-based system for optimization of fuel reload configurations

    International Nuclear Information System (INIS)

    Galperin, A.; Kimhi, S.; Segev, M.

    1989-01-01

    The authors discuss a knowledge-based production system developed for generating optimal fuel reload configurations. The system was based on a heuristic search method and implemented in Common Lisp programming language. The knowledge base embodied the reactor physics, reactor operations, and a general approach to fuel management strategy. The data base included a description of the physical system involved, i.e., the core geometry and fuel storage. The fifth cycle of the Three Mile Island Unit 1 pressurized water reactor was chosen as a test case. Application of the system to the test case revealed a self-learning process by which a relatively large number of near-optimal configurations were discovered. Several selected solutions were subjected to detailed analysis and demonstrated excellent performance. To summarize, applicability of the proposed heuristic search method in the domain of nuclear fuel management was proved unequivocally

  13. 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

  14. In-core fuel management: Reloading techniques. Proceedings of a technical committee meeting and workshop held in Vienna, 19-21 October 1992

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The purpose of the Technical Committee Meeting and Workshop on In-core Fuel Management - Reloading Techniques, convened by the IAEA in Vienna from 19 to 21 October 1992, was to provide an international forum to review and discuss in-core fuel management reloading techniques for light water reactors. A presentation of the history and status of reloading techniques was given by S.H. Levine, Pennsylvania State University, and papers on various computer code descriptions, methodologies and experiences of utilities and vendors for nuclear fuel reloading were presented and discussed. Optimization techniques for reloadings, expert system codes and the number of energy groups used in reloading calculations were discussed in more detail during a workshop session. Refs, figs and tabs.

  15. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    International Nuclear Information System (INIS)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia

    2017-01-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k ef f . The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi L ayer P erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  16. Azcaxalli: A system based on Ant Colony Optimization algorithms, applied to fuel reloads design in a Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel-Estrada, Jaime, E-mail: jaime.esquivel@fi.uaemex.m [Facultad de Ingenieria, Universidad Autonoma del Estado de Mexico, Cerro de Coatepec S/N, Toluca de Lerdo, Estado de Mexico 50000 (Mexico); Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Ortiz-Servin, Juan Jose, E-mail: juanjose.ortiz@inin.gob.m [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico); Castillo, Jose Alejandro; Perusquia, Raul [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico Toluca S/N, Ocoyoacac, Estado de Mexico 52750 (Mexico)

    2011-01-15

    This paper presents some results of the implementation of several optimization algorithms based on ant colonies, applied to the fuel reload design in a Boiling Water Reactor. The system called Azcaxalli is constructed with the following algorithms: Ant Colony System, Ant System, Best-Worst Ant System and MAX-MIN Ant System. Azcaxalli starts with a random fuel reload. Ants move into reactor core channels according to the State Transition Rule in order to select two fuel assemblies into a 1/8 part of the reactor core and change positions between them. This rule takes into account pheromone trails and acquired knowledge. Acquired knowledge is obtained from load cycle values of fuel assemblies. Azcaxalli claim is to work in order to maximize the cycle length taking into account several safety parameters. Azcaxalli's objective function involves thermal limits at the end of the cycle, cold shutdown margin at the beginning of the cycle and the neutron effective multiplication factor for a given cycle exposure. Those parameters are calculated by CM-PRESTO code. Through the Haling Principle is possible to calculate the end of the cycle. This system was applied to an equilibrium cycle of 18 months of Laguna Verde Nuclear Power Plant in Mexico. The results show that the system obtains fuel reloads with higher cycle lengths than the original fuel reload. Azcaxalli results are compared with genetic algorithms, tabu search and neural networks results.

  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. Artificial neural networks for spatial distribution of fuel assemblies in reload of PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Edyene; Castro, Victor F.; Velásquez, Carlos E.; Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Programa de Pós-Graduação em Ciências e Técnicas Nucleares

    2017-07-01

    An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reload. The main bounding parameter of the modelling was the neutron multiplication factor, k{sub ef{sub f}}. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. These parameters were obtained by the ORNL nuclear code package SCALE6.0. As for the artificial neural network, the ANN Feedforward Multi{sub L}ayer{sub P}erceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). Artificial neural network have implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core. (author)

  19. Report of lower endplug welding, and testing and inspecting result for MONJU 1{sup th} reload core fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Kajiyama, Takasi; Numata, Kazuaki; Ohtani, Seiji [Quality Assuranse Section, Technical Administration Division, Plutonium Fuel Center, Tokai Works, Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan); Kobayashi, Hiromi; Watanabe, Hiroaki; Goto, Tatsuro; Takahashi, Hideki; Nagasaku, Katsuhiko [Inspection Development Campany Ltd., Tokai, Ibaraki (Japan)

    2000-02-01

    The procedure and result of lower endplugwelding, Test and Inspection and Shipment of the 1{sup th} reload core fuel assembly (80 Fuel Assemblies) for the fast breeder reactor MONJU are reported, which had been examined and inspected in Tamatsukuri Branch, Material Insurance Office, Quality Assurance Section, Technical Administration Division, Plutonium Fuel Center (before: Inspection Section, Plutonium Fuel Division), from June 1994 to January 1996. The number of cladding tubes welded to the endplug were totally 13,804: 7,418 for Core - Inside of 43 fuel Assemblies and 6,836 for Core-Outside of 37 fuel Assemblies. 13,794 of them, 7,414 Core-Inside and 6,379 Core-Outside, were approved by the test and sent to Plutonium Fuel Center. 10 of them weren't approved mainly because of default welding. Disapproval rating was 0.07%. (author)

  20. Dry reloading and packaging of spent fuel at TRIGA MARK I reactor of Medical University Hanover (MHH), Germany

    International Nuclear Information System (INIS)

    Haferkamp, D.

    2008-01-01

    Between 1994 and 1998 the equipment for dry reloading of a research reactor was developed by Noell, which was funded by the German Federal Government and State of Saxonia. The task of this development programme was the design and delivery of an equipment able to load the spent fuel into the shipping casks in a dry mode for research reactors, where wet loading inside the storage pool is impossible. ALARA and infrastructure conditions had to be taken into consideration. Most of the research reactors of TRIGA MARK I type or WWR-SM have operating modes for handling of spent fuel inside the pond or for transfer of spent fuel from pond to dry/wet storage pools. On the other hand, most of them cannot handle heavy weighted shipping casks inside the reactor building because of the crane capacity, or inside water pool because of dimensions and weight of shipping casks. A typical licensed normal operating procedure for spent fuel in research reactors (TRIGA MARK I) is shown. Dry unloading procedure is described. Additionally to the normal operating procedures at the MHH research reactor the following steps were necessary: - dry packaging of spent fuel elements into the loading units (six packs) in order to minimise the transfer and loading steps between the pool and shipping cask; - transfer of spent fuel loading units from dry storage pool to the shipping cask (outside the reactor building) in a shielded transfer cask; - dry reloading of loading units, into the shipping casks outside the reactor building. The Dry Reloading Equipment implies the following 5 items: 1. loading units (six packs), which includes: - capacity up to six spent fuel elements; - criticality safe placement of spent fuel elements; - handling of several spent fuel elements in an aluminium loading unit. 2. Special Transfer Cask, which includes: - shielded housing with locks; - gripper inside housing; - hoist outside housing; - computer aided operation mode for loading and unloading. 3. Transfer Vehicle

  1. 11 th fuel reload of the Unit 1, leadership with results

    International Nuclear Information System (INIS)

    Serrano, R.H.

    2006-01-01

    At the moment the companies with world class, the formation of intellectual capital is a strategy to reach the excellence; the Management of Nucleo electric Centrals (GCN), aware of this strategy to achieve the mission commended, it establishes that it is decisive the leadership among their collaborators for the achievement of the goals. The obtained results in the 11 th reload of the Unit 1 are a sample of as how the leaders and collaborators when making work in team they have achieved the best results (collective dose and reload duration), until today in what is the history of the commercial operation of the Unit 1. (Author)

  2. Implement of MOX fuel assemblies in the design of the fuel reload for a BWR; Implemento de ensambles de combustible MOX en el diseno de la recarga de combustible para un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Alonso V, G.; Palacios H, J. C., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    At the present time the use of mixed oxides as nuclear fuel is a technology that has been implemented in mixed reloads of fuel for light water reactors. Due to the plutonium production in power reactors, is necessary to realize a study that presents the plutonium use like nuclear fuel. In this work a study is presented that has been carried out on the design of a fuel assembly with MOX to be proposed in the supply of a fuel reload. The fissile relationship of uranium to plutonium is presented for the design of the MOX assembly starting from plutonium recovered in the reprocessing of spent fuel and the comparison of the behavior of the infinite multiplication factor is presented and of the local power peak factor, parameters of great importance in the fuel assemblies design. The study object is a fuel assembly 10 x 10 GNF2 type for a boiling water reactor. The design of the fuel reload pattern giving fuel assemblies with MOX, so the comparison of the behavior of the stop margin for a fuel reload with UO{sub 2} and a mixed reload, implementing 12 and 16 fuel assemblies with MOX are presented. The results show that the implement of fuel assemblies with MOX in a BWR is possible, but this type of fuels creates new problems that are necessary to study with more detail. In the development of this work the calculus tools were the codes: INTREPIN-3, CASMO-4, CMSLINK and SIMULATE-3. (Author)

  3. Three stops of fuel reloading with length of less 30 days in the Laguna Verde Central

    International Nuclear Information System (INIS)

    Lozano L, A.

    2007-01-01

    The Laguna Verde Central having established as mission 'With maximum priority in the safety, to generate electricity by nuclear means with quality and cost competitive, sustained in our personnel's continuous overcoming and deep respect to the environment' and respecting our values (safety, responsibility by results, professional integrity, continuous improving, team working, excellence in the acting, quality of service, protection to the environment) they thought about our strategic objectives of the power station being born this way one of them that it is the program of improvement 'Reduction of reload times' looking for to be improves every day comparing us with the best plants in the world efficient all the processes in the power station that allowed us to measure our acting with the same parameters that settle down at international level like they are nuclear safety, industrial safety, radiological safety, capacity factor, readiness factor, cleaning of the power station attachment to procedures, attention to the detail and certainly to be competitive in the economic aspect. After analyzing the acting record of the power station, evaluating our technical capacity, economic, the location of the installation besides revising the international experiences it was defined that one of the concepts that impact considerably so much to the capacity and readiness factors besides the dose and production cost is the duration of the reload periods, for this reason they were elaborated work strategies to be able to reach our goal of reload days considered in being able to carry out them in less than 30 days, here the actions carried out that they made us complete the three last reloads in less than 30 days are captured. (Author)

  4. Fuel assemblies with inert matrices as reloads of cycle 11 of the Unit 1 of the LVNC; Ensamble combustibles con matrices inertes como recargas del ciclo 11 de la Unidad 1 de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez M, N.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2005-07-01

    In this work the results that were obtained of the analysis of three different reloads of the cycle 11 with fuel assemblies containing a mixture of UO{sub 2} and plutonium grade armament in an inert matrix. The proposed assemble, consists of an arrangement 10x10 with 42 bars fuels of PuO{sub 2}-CeO{sub 2}, 34 fuel bars with UO{sub 2} and 16 fuel bars with UO{sub 2}-Gd{sub 2O}3. The proposed assemble is equivalent to an it reloadable assemble of the cycle 11. The fuel bars of uranium and gadolinium, are of the same type of those that are used in the reloadable assemble of uranium. The design and generation of the nuclear databases of the fuel cell with mixed fuel, it was carried out with the HELIUMS code. The simulation of operation of the cycle 11, it was carried out with the CM-PRESTO code. The results show that with one reload of 72 assemblies of UO{sub 2} and 32 assemblies with mixed fuel has a cycle length of smaller in 10.5 days to the cycle length with the complete reload of assemblies of UO{sub 2} and a length smaller cycle in 34 days with the complete reload of 104 assemblies with mixed fuel. (Author)

  5. Optimization of fuel reloads for a BWR using the ant colony system; Optimizacion de recargas de combustible para un BWR usando el sistema de colonia de hormigas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J. [Universidad Autonoma del Estado de Mexico, Facultad de Ingenieria, Cerro de Coatepec s/n, Ciudad Universitaria, 50110 Toluca, Estado de Mexico (Mexico); Ortiz S, J. J. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jaime.es.jaime@gmail.com

    2009-10-15

    In this work some results obtained during the development of optimization systems are presented, which are employees for the fuel reload design in a BWR. The systems use the ant colony optimization technique. As first instance, a system is developed that was adapted at travel salesman problem applied for the 32 state capitals of Mexican Republic. The purpose of this implementation is that a similarity exists with the design of fuel reload, since the two problems are of combinatorial optimization with decision variables that have similarity between both. The system was coupled to simulator SIMULATE-3, obtaining good results when being applied to an operation cycle in equilibrium for reactors of nuclear power plant of Laguna Verde. (Author)

  6. Development of a new-generation system for reloading of nuclear fuel

    International Nuclear Information System (INIS)

    Maksimov, M.; Maslov, O.; Maisyan, I.

    1995-01-01

    The modern concept of development of nuclear power, which is also reflected in the new scientific and technical documentation, moves to the forefront the general systems aspects of performing operations with nuclear fuel. It is suggested that the organizational questions of delivering, accounting for, storing, monitoring, moving, calculating overloads and mechanisms, and devices which perform manipulation with nuclear fuel, be treated as a single system

  7. 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

  8. 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

  9. 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

  10. Four stops of fuel reloading with duration of less of 30 days in the Laguna Verde Central

    International Nuclear Information System (INIS)

    Lozano L, A.

    2008-01-01

    The Laguna Verde Central having established as mission 'With maximum priority in the safety, to generate electricity by nuclear means with competitive quality and cost, sustained in our personnel's continuous overcoming and deep respect to the environment' and respecting our values (safety, responsibility by results, professional integrity, continuous improvement, team work, excellence in the performance, quality of service, protection to the environment its thought about our strategic objectives of the power station being born by this way one of them that it is the improvement program 'reduction of reloading times' looking for to be improves every day comparing us with the best plants of the world effectiveness all the processes in the power station that allowed us to measure our performance with the same parameters that settle down at international level as its are nuclear safety, industrial safety, radiological safety, capacity factor, readiness factor, cleaning of the power station attachment to procedures, attention to the detail and certainly to be competitive in the economic aspect. After analyzing the performance record of the power station, evaluating our technical, economic capacity, the location of the installation besides revising the international experiences was defined that one of the concepts that impact considerably so much to the capacity factors and readiness besides the dose and production cost is the duration of the reloading periods, for this reason work strategies were elaborated to be able to reach our goals of reloading days in less than 30 days, here are formed the carried out actions that they made us complete the four last reloading in less than 30 days. (Author)

  11. Optimal reload and depletion method for pressurized water reactors

    International Nuclear Information System (INIS)

    Ahn, D.H.

    1984-01-01

    A new method has been developed to automatically reload and deplete a PWR so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: 1) determination of the minimum fuel requirement for an equivalent three-region core model, 2) optimal selection and allocation of fuel requirement for an equivalent three-region core model, 2) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the cost of the fresh reload fuel, 3) optimal placement of fuel assemblies to conserve regionwise optimal conditions and 4) optimal control through poison management to deplete individual fuel assemblies to maximize EOC k/sub eff/. Optimizing the fuel cost of reloading and depleting a PWR reactor cycle requires solutions to two separate optimization calculations. One of these minimizes the enriched fuel inventory in the core by optimizing the EOC k/sub eff/. The other minimizes the cost of the fresh reload cost. Both of these optimization calculations have now been combined to provide a new method for performing an automatic optimal reload of PWR's. The new method differs from previous methods in that the optimization process performs all tasks required to reload and deplete a PWR

  12. Reload shutdown for Nuclear Power Stations in spain in 2003

    International Nuclear Information System (INIS)

    2004-01-01

    Regarding time reductions in fuel reloading at Spanish nuclear power stations, the Spanish Nuclear Security Council (CSN), at the request of the Spanish Finance and Treasury Department of the Chamber of Deputies, delivered an instruction, by which power station's owners were urged to establish a detailed planning of reload operations. This article includes the results of this instruction. (Author) 6 refs

  13. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

    Havlůj, F.; Hejzlar, J.; Vočka, R.

    2013-01-01

    Performing core physics calculations for the sake of reload safety analysis is a very demanding and time consuming process. This process generally begins with the preparation of libraries for the core physics code using a lattice code. The next step involves creating a very large set of calculations with the core physics code. Lastly, the results of the calculations must be interpreted, correctly applying uncertainties and checking whether applicable limits are satisfied. Such a procedure requires three specialized experts. One must understand the lattice code in order to correctly calculate and interpret its results. The next expert must have a good understanding of the physics code in order to create libraries from the lattice code results and to correctly define all the calculations involved. The third expert must have a deep knowledge of the power plant and the reload safety analysis procedure in order to verify, that all the necessary calculations were performed. Such a procedure involves many steps and is very time consuming. At ÚJV Řež, a.s., we have developed a set of tools which can be used to automate and simplify the whole process of performing reload safety analysis. Our application QUADRIGA automates lattice code calculations for library preparation. It removes user interaction with the lattice code and reduces his task to defining fuel pin types, enrichments, assembly maps and operational parameters all through a very nice and user-friendly GUI. The second part in reload safety analysis calculations is done by CycleKit, a code which is linked with our core physics code ANDREA. Through CycleKit large sets of calculations with complicated interdependencies can be performed using simple and convenient notation. CycleKit automates the interaction with ANDREA, organizes all the calculations, collects the results, performs limit verification and displays the output in clickable html format. Using this set of tools for reload safety analysis simplifies

  14. Automatic optimization of a nuclear reactor reload using the algorithm Ant-Q

    International Nuclear Information System (INIS)

    Machado, Liana; Schirru, Roberto

    2002-01-01

    The nuclear fuel reload optimization is a NP-Complete combinatorial optimization problem. For decades this problem was solved using an expert's knowledge. From the eighties, however there have been efforts to automatic fuel reload and the more recent ones show the Genetic Algorithm's (GA) efficiency on this problem. Following this trend, our aim is to optimization nuclear fuel reload using Ant-Q, artificial theory based algorithms. Ant-Q's results on the Traveling salesman Problem, which is conceptuality similar to fuel reload, are better than GA's. Ant-Q was tested in real application on the cycle 7 reload of Angra I. Comparing Ant-Q result with the GA's, it can be verified that, even without a local heuristics, the former algorithm, as it superiority comparing the GA in Angra I show. Is a valid technique to solve the nuclear fuel reload problem. (author)

  15. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  16. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  17. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maddock, Thomas L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Ning [Idaho National Lab. (INL), Idaho Falls, ID (United States); Phillips, Ann Marie [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schreck, Kenneth A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolin, John M. [General Atomics, San Diego, CA (United States); Veca, Anthony [General Atomics, San Diego, CA (United States); McKnight, Richard D. [Argonne National Lab. (ANL), Argonne, IL (United States); Lell, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  18. Safety analysis and optimization of the core fuel reloading for the Moroccan TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Nacir, B.; Boulaich, Y.; Chakir, E.; El Bardouni, T.; El Bakkari, B.; El Younoussi, C.

    2014-01-01

    Highlights: • Additional fresh fuel elements must be added to the reactor core. • TRIGA reactor could safely operate around 2 MW power with 12% fuel elements. • Thermal–hydraulic parameters were calculated and the safety margins are respected. • The 12% fuel elements will have no influence on the safety of the reactor. - Abstract: The Moroccan TRIGA MARK II reactor core is loaded with 8.5% in weight of uranium standard fuel elements. Additional fresh fuel elements must periodically be added to the core in order to remedy the observed low power and to return to the initial reactivity excess at the End Of Cycle. 12%-uranium fuel elements are available to relatively improve the short fuel lifetime associated with standard TRIGA elements. These elements have the same dimensions as standards elements, but with different uranium weight. The objective in this study is to demonstrate that the Moroccan TRIGA reactor could safely operate, around 2 MW power, with new configurations containing these 12% fuel elements. For this purpose, different safety related thermal–hydraulic parameters have been calculated in order to ensure that the safety margins are largely respected. Therefore, the PARET model for this TRIGA reactor that was previously developed and combined with the MCNP transport code in order to calculate the 3-D temperature distribution in the core and all the most important parameters like the axial distribution of DNBR (Departure from Nucleate Boiling Ratio) across the hottest channel. The most important conclusion is that the 12% fuel elements utilization will have no influence on the safety of the reactor while working around 2 MW power especially for configurations based on insertions in C and D-rings

  19. Immersed multiple device for the control of the irradiated PWR fuel pins in the reloadable loop in the OSIRIS pond

    International Nuclear Information System (INIS)

    Farny, G.

    1983-01-01

    With respect to the dynamics of the degradation of the PWR fuel in transient, normal and abnormal regions, a new multi-device immersed in the cooling pond of the OSIRIS reactor, is studied. The multiple device is subjected to three examinations: (1) visual studying and video-recording of the appearance of the fuel pins, (2) metrology of the pins, (3) investigation of the induced Foucault currents in the fuel cans. Attention is chiefly paid to the last point; the other ones - being closely related - are only touched on whenever needed. It is concluded that quality control of the fuel pins is possible by means of Foucault currents without applying mechanical constraints and without interfering with the cooling rate. (Auth.)

  20. A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading

    International Nuclear Information System (INIS)

    Binh, Do Quang; Huy, Ngo Quang; Hai, Nguyen Hoang

    2014-01-01

    This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

  1. A binary mixed integer coded genetic algorithm for multi-objective optimization of nuclear research reactor fuel reloading

    Energy Technology Data Exchange (ETDEWEB)

    Binh, Do Quang [University of Technical Education Ho Chi Minh City (Viet Nam); Huy, Ngo Quang [University of Industry Ho Chi Minh City (Viet Nam); Hai, Nguyen Hoang [Centre for Research and Development of Radiation Technology, Ho Chi Minh City (Viet Nam)

    2014-12-15

    This paper presents a new approach based on a binary mixed integer coded genetic algorithm in conjunction with the weighted sum method for multi-objective optimization of fuel loading patterns for nuclear research reactors. The proposed genetic algorithm works with two types of chromosomes: binary and integer chromosomes, and consists of two types of genetic operators: one working on binary chromosomes and the other working on integer chromosomes. The algorithm automatically searches for the most suitable weighting factors of the weighting function and the optimal fuel loading patterns in the search process. Illustrative calculations are implemented for a research reactor type TRIGA MARK II loaded with the Russian VVR-M2 fuels. Results show that the proposed genetic algorithm can successfully search for both the best weighting factors and a set of approximate optimal loading patterns that maximize the effective multiplication factor and minimize the power peaking factor while satisfying operational and safety constraints for the research reactor.

  2. Improvement in PWR automatic optimization reloading methods using genetic algorithm

    International Nuclear Information System (INIS)

    Levine, S.H.; Ivanov, K.; Feltus, M.

    1996-01-01

    The objective of using automatic optimized reloading methods is to provide the Nuclear Engineer with an efficient method for reloading a nuclear reactor which results in superior core configurations that minimize fuel costs. Previous methods developed by Levine et al required a large effort to develop the initial core loading using a priority loading scheme. Subsequent modifications to this core configuration were made using expert rules to produce the final core design. Improvements in this technique have been made by using a genetic algorithm to produce improved core reload designs for PWRs more efficiently (authors)

  3. Improvement in PWR automatic optimization reloading methods using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Levine, S H; Ivanov, K; Feltus, M [Pennsylvania State Univ., University Park, PA (United States)

    1996-12-01

    The objective of using automatic optimized reloading methods is to provide the Nuclear Engineer with an efficient method for reloading a nuclear reactor which results in superior core configurations that minimize fuel costs. Previous methods developed by Levine et al required a large effort to develop the initial core loading using a priority loading scheme. Subsequent modifications to this core configuration were made using expert rules to produce the final core design. Improvements in this technique have been made by using a genetic algorithm to produce improved core reload designs for PWRs more efficiently (authors).

  4. TVA experience in BWR reload design and licensing

    International Nuclear Information System (INIS)

    Robertson, J.D.

    1986-01-01

    TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

  5. 11 th fuel reload of the Unit 1, leadership with results; 11a recarga de combustible de la Unidad 1, liderazgo con resultados

    Energy Technology Data Exchange (ETDEWEB)

    Serrano, R.H. [Comision Federal de Electricidad, Mexico D.F. (Mexico)]. e-mail: hsr98581@cfe.gob.mx

    2006-07-01

    At the moment the companies with world class, the formation of intellectual capital is a strategy to reach the excellence; the Management of Nucleo electric Centrals (GCN), aware of this strategy to achieve the mission commended, it establishes that it is decisive the leadership among their collaborators for the achievement of the goals. The obtained results in the 11 th reload of the Unit 1 are a sample of as how the leaders and collaborators when making work in team they have achieved the best results (collective dose and reload duration), until today in what is the history of the commercial operation of the Unit 1. (Author)

  6. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors

    International Nuclear Information System (INIS)

    Ortega C, R.F.

    2008-01-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U 3O 8 in January, 2005 to a maximum of US$137.00 dollars by Ib U 3 O 8 by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U 3 O 8 such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  7. Cuckoo Search with flight of Levy applied to the problem of reload of fuels in nuclear reactors

    International Nuclear Information System (INIS)

    Silva, Patrick V.; Nast, Fernando N.; Schirru, Roberto; Meneses, Anderson A.M.; Coordenacao de Pos-Graduacao e Pesquisa de Engenharia

    2017-01-01

    Intra-Nuclear Fuel Management Optimization is a complex combinatorial problem of the NP-difficult type, associated with the refueling process of a nuclear reactor, which aims to extend the cycle of operation by determining loading patterns, obeying safety margins. In addition to the combinatorial problem, we have the aspect of calculations of reactor physics, which increases the difficult of OGCIN. Methods that are proving effective when applied to OGCIN are the algorithms belonging to the swarm intelligence paradigm. A new member of this paradigm is Cuckoo Search (CS), which has shown results promising when applied to optimization issues. The CS is based on the litter parasitism of some cuckoo species combined with the Levy flight behavior of some birds. In the present work we present the results of the application of CS to OGCIN, and compare them to the results obtained by the application of ABC

  8. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  9. Reload core safety verification

    International Nuclear Information System (INIS)

    Svetlik, M.; Minarcin, M.

    2003-01-01

    This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

  10. Contribution to the methodology of safety evaluation - and licensing of reloading cycle for PWR type reactors

    International Nuclear Information System (INIS)

    Esteves, R.G.

    1981-01-01

    A simplified methodology for evaluating a reload safety cycle is presented. This methodology consists in selecting for each foreseen accident, the nuclear key reload safety parameters which determine the accident evolution. So, each key reload parameter is calculated and compared with its value for the first cycle. Those accidents, which have their key reload parameter bounded by the values of the first cycle do not need reanalise. Extension of the validity of this methodology when there exists change of fuel supplier is commented. (Author) [pt

  11. Study of heuristics in ant system for nuclear reload optimisation

    International Nuclear Information System (INIS)

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

    2007-01-01

    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 loading pattern that maximizes the number of effective full power days, minimizing the relationship cost/benefit. 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 used to solve the nuclear reactor core fuel reload optimization problem, with compatibles heuristics. 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)

  12. Study of heuristics in ant system for nuclear reload optimisation

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Schirru, Roberto; Silva, Fernando C. da; Machado, Marcelo D.; Medeiros, Jose A.C.C. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mail: alan@lmp.ufrj.br; schirru@lmp.ufrj.br; fernando@con.ufrj.br; marcelo@lmp.ufrj.br; canedo@lmp.ufrj.br

    2007-07-01

    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 loading pattern that maximizes the number of effective full power days, minimizing the relationship cost/benefit. 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 used to solve the nuclear reactor core fuel reload optimization problem, with compatibles heuristics. 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)

  13. Reloadable radioactive generator system

    International Nuclear Information System (INIS)

    Colombetti, L.G.

    1977-01-01

    A generator system that can be reloaded with an elutable radioactive material, such as 99 molybdenum, a multiple number of times is described. The system basically comprises a column filled with alumina, a loading vial containing a predetermined amount of the elutable radioactive material, and a rinsing vial containing a sterile solution. The two vials are connected by a conduit so that when communication is achieved between the column and loading vial and an evacuated vial is placed in communication with the bottom of the column, the predetermined amount of the radioactive material in the loading vial will be transferred to the column. The procedure can be repeated as the elutable material in the column is dissipated

  14. Parallel genetic algorithm as a tool for nuclear reactors reload

    International Nuclear Information System (INIS)

    Santos, Darley Roberto G.; Schirru, Roberto

    1999-01-01

    This work intends to present a tool which can be used by designers in order to get better solutions, in terms of computational costs, to solve problems of nuclear reactor reloads. It is known that the project of nuclear fuel reload is a complex combinatorial one. Generally, iterative processes are the most used ones because they generate answers to satisfy all restrictions. The model presented here uses Artificial Intelligence techniques, more precisely Genetic Algorithms techniques, mixed with parallelization techniques.Test of the tool presented here were highly satisfactory, due to a considerable reduction in computational time. (author)

  15. Automatic optimization of a nuclear reactor reload using the algorithm Ant-Q; A otimizacao automatica da recarga nuclear utilizando o algoritmo Ant-Q

    Energy Technology Data Exchange (ETDEWEB)

    Machado, Liana; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The nuclear fuel reload optimization is a NP-Complete combinatorial optimization problem. For decades this problem was solved using an expert's knowledge. From the eighties, however there have been efforts to automatic fuel reload and the more recent ones show the Genetic Algorithm's (GA) efficiency on this problem. Following this trend, our aim is to optimization nuclear fuel reload using Ant-Q, artificial theory based algorithms. Ant-Q's results on the Traveling salesman Problem, which is conceptuality similar to fuel reload, are better than GA's. Ant-Q was tested in real application on the cycle 7 reload of Angra I. Comparing Ant-Q result with the GA's, it can be verified that, even without a local heuristics, the former algorithm, as it superiority comparing the GA in Angra I show. Is a valid technique to solve the nuclear fuel reload problem. (author)

  16. Reload safety evaluation report for Ulchin nuclear power plant unit 2, cycle 6

    International Nuclear Information System (INIS)

    Chung, Jin Gon; Park, Jin Ha; Kim, Oh Hwan; Oh, Dong Seok; Kim, Du Ill; Choi, Han Rim; Ku, Dong Uk; Bae, Hoo Gun

    1994-07-01

    This report presents a reload safety evaluation for Ulchin-2, cycle 6 and demonstrates that the core being composed of various fuel types as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 6 core and results are described in the report. (Author) 1 ref., 3 figs., 7 tabs

  17. Reload safety evaluation report for ulchin nuclear power plant unit 2, cycle 4

    International Nuclear Information System (INIS)

    Park, Chan Oh; Park, Yong Soo; Kim, Hong Jin; Kim, Il Kon; Oh, Dong Seok; Yoon, Han Yong; Choi, Han Rim; Choi, Dong Uk; Lee, Chung Chan; Zee, Sung Kyun

    1992-09-01

    This report presents a reload safety evaluation for Ulchin-2, Cycle 4 and demonstrates that the core being composed of various fuel types as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 core and results are described in the report. (Author)

  18. Reload safety evaluation report for Kori nuclear power plant unit 1, cycle 13

    International Nuclear Information System (INIS)

    Park, Chan Oh; Moon, Bok Ja; Cho, Byeong Ho; Nam, Kee Il; Kim, Oh Hwan; Chang, Doo Soo; Yoon, Han Young; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

    1993-03-01

    This report presents the reload safety evaluation for Kori-1, Cycle 13 and demonstrates that the reactor core being composed of various fuel assembly types applied in this evaluation will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 13 core and results are described in this report. (Author)

  19. Reload safety evaluation report for Ulchin nuclear power plant unit 1 cycle 5

    International Nuclear Information System (INIS)

    Park, Chan Oh; Kim, Yong Rae; Son, Sang Rin; Oh, Dong Seok; Kim, Hong Jin; Yoon, Kyung Ho; Yoon, Han Young; Choi, Han Rim; Choi, Dong Uk

    1992-12-01

    This report presents the reload safety evaluation for Ulchin 1, Cycle 5 and demonstrates that the reactor core being composed of various fuel assembly types applied in this evaluation will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 5 core and results are described in this report. (Author)

  20. The high moderating ratio reactor using 100% MOX reloads

    International Nuclear Information System (INIS)

    Barbrault, P.

    1994-06-01

    This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

  1. PWR reload safety evaluation methodology

    International Nuclear Information System (INIS)

    Doshi, P.K.; Chapin, D.L.; Love, D.S.

    1993-01-01

    The current practice for WWER safety analysis is to prepare the plant Safety Analysis Report (SAR) for initial plant operation. However, the existing safety analysis is typically not evaluated for reload cycles to confirm that all safety limits are met. In addition, there is no systematic reanalysis or reevaluation of the safety analyses after there have been changes made to the plant. The Westinghouse process is discussed which is in contrast to this and in which the SAR conclusions are re-validated through evaluation and/or analysis of each reload cycle. (Z.S.)

  2. Improvement of characteristic statistic algorithm and its application on equilibrium cycle reloading optimization

    International Nuclear Information System (INIS)

    Hu, Y.; Liu, Z.; Shi, X.; Wang, B.

    2006-01-01

    A brief introduction of characteristic statistic algorithm (CSA) is given in the paper, which is a new global optimization algorithm to solve the problem of PWR in-core fuel management optimization. CSA is modified by the adoption of back propagation neural network and fast local adjustment. Then the modified CSA is applied to PWR Equilibrium Cycle Reloading Optimization, and the corresponding optimization code of CSA-DYW is developed. CSA-DYW is used to optimize the equilibrium cycle of 18 month reloading of Daya bay nuclear plant Unit 1 reactor. The results show that CSA-DYW has high efficiency and good global performance on PWR Equilibrium Cycle Reloading Optimization. (authors)

  3. Automatic optimized reload and depletion method for a pressurized water reactor

    International Nuclear Information System (INIS)

    Ahn, D.H.; Levene, S.H.

    1985-01-01

    A new method has been developed to automatically reload and deplete a pressurized water reactor (PWR) so that both the enriched inventory requirements during the reactor cycle and the cost of reloading the core are minimized. This is achieved through four stepwise optimization calculations: (a) determination of the minimum fuel requirement for an equivalent three-region core model, (b) optimal selection and allocation of fuel assemblies for each of the three regions to minimize the reload cost, (c) optimal placement of fuel assemblies to conserve regionwise optimal conditions, and (d) optimal control through poison management to deplete individual fuel assemblies to maximize end-of-cycle k /SUB eff/ . The new method differs from previous methods in that the optimization process automatically performs all tasks required to reload and deplete a PWR. In addition, the previous work that developed optimization methods principally for the initial reactor cycle was modified to handle subsequent cycles with fuel assemblies having burnup at beginning of cycle. Application of the method to the fourth reactor cycle at Three Mile Island Unit 1 has shown that both the enrichment and the number of fresh reload fuel assemblies can be decreased and fully amortized fuel assemblies can be reused to minimize the fuel cost of the reactor

  4. 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

  5. Economical analysis of the second partial reload for Angra 1 with partial low-leakage

    International Nuclear Information System (INIS)

    Mascarenhas, H.A.; Teixeira, M.C.C.; Dias, A.M.

    1990-01-01

    Preliminary results for the Angra 1 second reload design with partial low-leakage were assessed with NUCOST 1.0, code for nuclear power costs calculation. In the proposed scheme, some partially burned fuel assemblies (FAs) are located at the core boundary, while new FAs occupy more internal positions. The nuclear design - utilizing the code system SAV (from Siemens/KWU Group, F.R. Germany) - has been performed with detail for the 3rd cycle while simpler approach has been utilized for subsequent reloads. Results of NUCOST 1.0 show that the partial low-leakage reload in the 3rd cycle of Angra 1 offers fuel costs 1% lower when compared to the Plant's actual reload scheme, what corresponds to an savings of about US$190.000. When operation and maintenance and capital costs are also considered, economies in the order of US$2.6 million are obrained. (author) [pt

  6. 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

  7. Reload safety evaluation report for Ulchin nuclear power plant unit 1, cycle 6

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Yong Rae; Kim, Oh Hwan; Kwon, Hyuk Sung; Yoon, Han Young; Choi, Han Rim; Ku, Dong Uk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    This report presents a reload safety evaluation for Ulchin 1, cycle 6 and demonstrates that the reactor core being fully composed of KOFA as described in this report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 6 core and results are described in this report. (Author) 1 ref., 5 figs., 6 tabs.

  8. Reload safety evaluation report for Ulchin nuclear power plant unit 1, cycle 6

    International Nuclear Information System (INIS)

    Lee, Won Jae; Kim, Yong Rae; Kim, Oh Hwan; Kwon, Hyuk Sung; Yoon, Han Young; Choi, Han Rim; Ku, Dong Uk

    1993-11-01

    This report presents a reload safety evaluation for Ulchin 1, cycle 6 and demonstrates that the reactor core being fully composed of KOFA as described in this report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 6 core and results are described in this report. (Author) 1 ref., 5 figs., 6 tabs

  9. Reload safety evaluation report for Yonggwang nuclear power plant unit 1, cycle 8

    International Nuclear Information System (INIS)

    Lee, Won Jae; Yoon, Kyung Ho; Cho, Young Chul; Kim, Jae Hak; Um, Kil Sup; Choi, Han Rim; Kim, Ki Hang; Sung, Kang Sik

    1993-09-01

    This report presents a reload safety evaluation for YGN-1, cycle 8 and demonstrates that the core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the cycle 8 core and results are described in the report. (Author) 1 ref., 4 figs., 5 tabs

  10. Reload safety evaluation report for yonggwang nuclear power plant unit 2 cycle 7

    International Nuclear Information System (INIS)

    Zee, Sung Kyun; Choi, Gyoo Hwan; Lee, Ki Bog; Park, Sang Yoon

    1993-01-01

    This report presents the reload safety evaluation for YGN-2, Cycle 7 and demonstrates that the reactor core being entirely composed of KOFA as described below will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which would potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 7 core design described herein. (Author)

  11. Reload safety evaluation report for yonggwang nuclear power plant unit 1 cycle 7

    International Nuclear Information System (INIS)

    Park, Chan Oh; Kwon, Tae Je; Park, Sang Yoon; Sung, Kang Sik; Kim, Ki Hang; Yim, Jeong Sik; Kim, Du Ill; Choi, Han Rim; Bae, Hoo Gun

    1992-06-01

    This report presents the reload safety evaluation for YGN-1, Cycle 7 and demonstrates that the reactor core being entirely composed of KOFA as discribed below will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the reload fuel assemblies have been reviewed for the Cycle 7 core and results are described in this report. (Author)

  12. Reload safety evaluation report for Yonggwang nuclear power plant unit 1, cycle 9

    International Nuclear Information System (INIS)

    Cho, Young Chul; Nam, Kee Il; Kim, Ki Hang; Suh, Jung Min; Um, Kil Sup; Ban, Chang Hwan; Bae, Hoo Gun

    1995-02-01

    This report presents a reload safety evaluation for YGN-1, Cycle 9 and demonstrates that the core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 9 core and results are described in this report. (Author) 1 refs., 3 figs., 6 tabs

  13. Reload safety evaluation report for Ulchin nuclear power plant unit 1, cycle 7

    International Nuclear Information System (INIS)

    Kim, Yong Rae; Kwon, Hyuk Sung; Kim, Oh Hwan; Choi, Han Rim; Yoon, Han Young; Ku, Dong Uk; Suh, Jung Min; Bae, Hoo Gun

    1995-02-01

    This report presents a reload safety evaluation for UCN-2, Cycle 7 and demonstrates that the core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 7 core and results are described in this report. (Author) 1 refs., 3 figs., 6 tabs

  14. Reload safety evaluation report for Kori nuclear power unit 1, cycle 14

    International Nuclear Information System (INIS)

    Kim, Joo Young; Kim, Oh Hwan; Nam, Kee Il; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

    1994-05-01

    This report presents the reload safety evaluation for Kori-1, Cycle 14 and demonstrate that the reactor core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 14 core design described herein. (Author) 1 refs., 9 figs., 5 tabs

  15. Reload safety evaluation report for kori nuclear power plant unit 4, cycle 8

    International Nuclear Information System (INIS)

    Park, Chan Oh; Jung, Yil Sup; Kim, Si Yong; Kim, Ki Hang; Kwon, Hyuk Sung; Oh, Dong Seok; Kim, Du Ill; Ban, Chang Hwan; Choi, Dong Uk

    1993-06-01

    This report presents the reload safety evaluation for Kori-4, Cycle 8 and demonstrate that the reactor core being entirely composed of KOFA as described in the report will not adversely affect the safety of the public and the plant. All of the accidents comprising the licening bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 8 core design described herein. (Author)

  16. Historical event: the best reload of CLV

    International Nuclear Information System (INIS)

    Rivera C, A.

    2006-01-01

    The present work will describe how the Laguna Verde Central advances to achieve the excellence in the evaluations of WANO, in one of their two important concepts that have to improve to aspire to that qualification, in Collective Dose of the Personnel and in the duration of the periods of the fuel reloads. The result of the 11th Fuel reload of the Unit 1 that begins September 11, 2005 and concludes on October 10, 2005, came out in 29.5 days, being the best in the history of the CLV. It critical route (most large duration activities) it was the bigger maintenance of the main generator, having like challenge that if the program was completed or decreased the time, the Laguna Verde Central it would be placed in the threshold of the qualification of Excellency. With this vision the technicians of the Power station in previous meetings, determined to improve the maneuvers recommended by the manufacturing Mitsubishi (Japanese Company) for the disassembly and assembly of the rotor that it has a weight of 120 tons, same that for their great weight make difficult the maneuver. This maintenance is carried out every five years, highlighting that in its previous maintenance it was in reload of 17 hours in its retirement, same that it was to improve. The changes to the instructions of Mitsubishi for the retirement of the rotor are: the use of synthetic slings instead of steel strobes that needed to use of neoprene and wooden staves to protect the rotor and a system of pulleys instead of differential of chain, being 3 hours in the retirement of the rotor, instead of 17 hours that it lasted previously. It is also excellent that it decreased of six necessary people for the movement of the rotor, to a one person. The reduction of effort was shown in pictures, like instead of the chain differential, they put on steel strobes with pulleys. To give the force of the movement it was used a hoisting machine. The two fixed support points were changed in supports with simple pulleys

  17. Reload safety evaluation report for kori nuclear power plant unit 2 cycle 9

    International Nuclear Information System (INIS)

    Cho, Beom Jin; Kim, Si Yong; Kim, Oh Hwan; Nam, Kee Il; Um, Gil Sup; Ban, Chang Hwan; Choi, Dong Uk; Yoon, Kyung Ho

    1992-04-01

    The Kori Nuclear Power Plant Unit 2 (Kori-2) is anticipated to be refuelled with 16x16 Korean Fuel Assemblies (KOFA), which are based on the KAERI design starting from Cycle 8. This report presents a reload safety evaluation for Kori-2, Cycle 9 and demonstrates that the reactor core being composed of various fuel assembly types as described below will not adversely affect the safety of the public and the plant. The evaluation of Kori-2, Cycle 9 was accomplished utilizing the methodology described in 'Reload Transition Safety Report for KORI 2' (Ref. /1-1/). The reload core for Kori-2, Cycle 9 is entirely comprised of 16x16 KOFA. In the Kori-2 licensing documentation to KEPCO the reference safety evaluation was provided for the operation of a reactor core fully loaded with KOFA as well as associated proposed changes to the Kori-2 Technical Specifications. The reload for Kori-2, Cycle 9 also introduces UO 2 /Gd 2 O 3 containing fuel rods. The use of fuel rods with Gd 2 O 3 poisoning of the fuel has been approved as a part of the above mentioned licensing documentation. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 9 core design described herein. (Author)

  18. A nuclear reload optimization approach using a real coded genetic algorithm with random keys

    International Nuclear Information System (INIS)

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

    2009-01-01

    The fuel reload of a Pressurized Water Reactor is made whenever the burn up of the fuel assemblies in the nucleus of the reactor reaches a certain value such that it is not more possible to maintain a critical reactor producing energy at nominal power. The problem of fuel reload optimization consists on determining the positioning of the fuel assemblies within the nucleus of the reactor in an optimized way to minimize the cost benefit relationship of fuel assemblies cost per maximum burn up, and also satisfying symmetry and safety restrictions. The fuel reload optimization problem difficulty grows exponentially with the number of fuel assemblies in the nucleus of the reactor. During decades the fuel reload optimization problem was solved manually by experts that used their knowledge and experience to build configurations of the reactor nucleus, and testing them to verify if safety restrictions of the plant are satisfied. To reduce this burden, several optimization techniques have been used, included the binary code genetic algorithm. In this work we show the use of a real valued coded approach of the genetic algorithm, with different recombination methods, together with a transformation mechanism called random keys, to transform the real values of the genes of each chromosome in a combination of discrete fuel assemblies for evaluation of the reload optimization. Four different recombination methods were tested: discrete recombination, intermediate recombination, linear recombination and extended linear recombination. For each of the 4 recombination methods 10 different tests using different seeds for the random number generator were conducted 10 generating, totaling 40 tests. The results of the application of the genetic algorithm are shown with formulation of real numbers for the problem of the nuclear reload of the plant Angra 1 type PWR. Since the best results in the literature for this problem were found by the parallel PSO we will it use for comparison

  19. Spent fuel bundle counter sequence error manual - BRUCE NGS

    International Nuclear Information System (INIS)

    Nicholson, L.E.

    1992-01-01

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  20. Spent fuel bundle counter sequence error manual - DARLINGTON NGS

    International Nuclear Information System (INIS)

    Nicholson, L.E.

    1992-01-01

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  1. Hybrid expert system implementation to determine core reload patterns

    International Nuclear Information System (INIS)

    Greek, K.J.; Robinson, A.H.

    1989-01-01

    Determining reactor reload fuel patterns is a computationally intensive problem solving process for which automation can be of significant benefit. Often much effort is expended in the search for an optimal loading. While any modern programming language could be used to automate solution, the specialized tools of artificial intelligence (AI) are the most efficient means of introducing the fuel management expert's knowledge into the search for an optimum reload pattern. Prior research in pressurized water reactor refueling strategies developed FORTRAN programs that automated an expert's basic knowledge to direct a search for an acceptable minimum peak power loading. The dissatisfaction with maintenance of compiled knowledge in FORTRAN programs has served as the motivation for the development of the SHUFFLE expert system. SHUFFLE is written in Smalltalk, an object-oriented programming language, and evaluates loadings as it generates them using a two-group, two-dimensional nodal power calculation compiled in a personal computer-based FORTRAN. This paper reviews the object-oriented representation developed to solve the core reload problem with an expert system tool and its operating prototype, SHUFFLE

  2. Reload pattern optimization by application of multiple cyclic interchange algorithms

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Quist, A.J.; Hoogenboom, J.E. [Technische Univ. Delft (Netherlands)

    1996-09-01

    Reload pattern optimization procedures are proposed which are based on the multiple cyclic interchange approach, according to which the search for the reload pattern associated with the highest objective function value can be thought of as divided in multiple stages. The transition from the initial to the final stage is characterized by an increase in the degree of locality of the search procedure. The general idea is that, during the first stages, the `elite` cluster containing the group of best patterns must be located, after which the solution space is sampled in a more and more local sense to find the local optimum in this cluster. The transition(s) from global search behaviour to local search behaviour can be either prompt, by defining strictly separate search regimes, or gradual by introducing stochastic tests for the number of fuel bundles involved in a cyclic interchange. Equilibrium cycle optimization results are reported for a test PWR reactor core of modest size. (author)

  3. Reload pattern optimization by application of multiple cyclic interchange algorithms

    International Nuclear Information System (INIS)

    Geemert, R. van; Quist, A.J.; Hoogenboom, J.E.

    1996-01-01

    Reload pattern optimization procedures are proposed which are based on the multiple cyclic interchange approach, according to which the search for the reload pattern associated with the highest objective function value can be thought of as divided in multiple stages. The transition from the initial to the final stage is characterized by an increase in the degree of locality of the search procedure. The general idea is that, during the first stages, the 'elite' cluster containing the group of best patterns must be located, after which the solution space is sampled in a more and more local sense to find the local optimum in this cluster. The transition(s) from global search behaviour to local search behaviour can be either prompt, by defining strictly separate search regimes, or gradual by introducing stochastic tests for the number of fuel bundles involved in a cyclic interchange. Equilibrium cycle optimization results are reported for a test PWR reactor core of modest size. (author)

  4. Reactivity monitoring during reactor-reloading operations

    International Nuclear Information System (INIS)

    Baumann, N.P.; Ahlfeld, C.F.; Ridgely, G.C.

    1983-01-01

    At the Savannah River Plant (SRP) reloading operations during shutdown present special considerations in reactivity monitoring and control. Large reactivity changes may occur during reloading operations because of the heterogeneous nature of some core designs. This paper describes an improved monitoring system

  5. Controlling Oesophageal Variceal Bleeding by Reloading ...

    African Journals Online (AJOL)

    A special reloading kit (produced by McGown; USA) was used to reload previously used and sterilized Opti-vu caps from Saeed six shooter variceal band ligators (North Carolina, USA). Subjects with oesophageal varices underwent banding of the varices down the lower 5cm of the oesophagus using this technique.

  6. A boolean optimization method for reloading a nuclear reactor

    International Nuclear Information System (INIS)

    Misse Nseke, Theophile.

    1982-04-01

    We attempt to solve the problem of optimal reloading of fuel assemblies in a PWR, without any assumption on the fuel nature. Any loading is marked by n 2 boolean variables usub(ij). The state of the reactor is characterized by his Ksub(eff) and the related power distribution. The resulting non-linear allocation problems are solved throught mathematical programming technics combining the simplex algorithm and an extension of the Balas-Geoffrion's one. Some optimal solutions are given for PWR with assemblies of different enrichment [fr

  7. Development of the RSAC Automation System for Reload Core of WH NPP

    International Nuclear Information System (INIS)

    Choi, Yu Sun; Bae, Sung Man; Koh, Byung Marn; Hong, Sun Kwan

    2006-01-01

    The Nuclear Design for Reload Core of Westinghouse Nuclear Power Plant consists of 'Reload Core Model Search', 'Safety Analysis(RSAC)', 'NDR(Nuclear Design Report) and OCAP(Operational Core Analysis Package Generation)' phases. Since scores of calculations for various accidents are required to confirm that the safety analysis assumptions are valid, the Safety Analysis(RSAC) is the most important and time and effort consuming phase of reload core design sequence. The Safety Analysis Automation System supports core designer by the automation of safety analysis calculations in 'Safety Analysis' phase(about 20 calculations). More than 10 kinds of codes, APA(ALPHA/PHOENIX/ANC), APOLLO, VENUS, PHIRE XEFIT, INCORE, etc. are being used for Safety Analysis calculations. Westinghouse code system needs numerous inputs and outputs, so the possibility of human errors could not be ignored during Safety Analysis calculations. To remove these inefficiencies, all input files for Safety Analysis calculations are automatically generated and executed by this Safety Analysis Automation System. All calculation notes are generated and the calculation results are summarized in RSAC (Reload Safety Analysis Checklist) by this system. Therefore, The Safety Analysis Automation System helps the reload core designer to perform safety analysis of the reload core model instantly and correctly

  8. Reverse depletion method for PWR core reload design

    International Nuclear Information System (INIS)

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

    1985-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

  9. Three-batch reloading scheme for IRIS reactor extended cycles

    International Nuclear Information System (INIS)

    Jecmenica, R.; Pevec, D.; Grgic, D.

    2004-01-01

    To fully exploit the IRIS reactor optimized maintenance, and at the same time improve fuel utilization, a core design enabling a 4-year operating cycle together with a three-batch reloading scheme is desirable. However, this requires not only the increased allowed burnup but also use of fuel with uranium oxide enriched beyond 5%. This paper considers three-batch reloading scheme for a 4-year operating cycle with the assumptions of increased discharge burnup and fuel enrichment beyond 5%. Calculational model of IRIS reactor core has been developed based on FER FA2D code for group constants generation and NRC's PARCS nodal code for global core analysis. Studies have been performed resulting in a preliminary design of a three-batch core configuration for the first cycle. It must be emphasized that this study is outside the current IRIS licensing efforts, which rely on the present fuel technology (enrichment below 5%), but it is of long-term interest for potential future IRIS design upgrades. (author)

  10. Four stops of fuel reloading with duration of less of 30 days in the Laguna Verde Central; Cuatro paradas de recarga de combustible con duracion de menos de 30 dias en la central laguna verde

    Energy Technology Data Exchange (ETDEWEB)

    Lozano L, A. [CFE, Central Laguna Verde, Planeacion, Veracruz (Mexico)]. e-mail: agustin.lozano@cfe.gob.mx

    2008-07-01

    The Laguna Verde Central having established as mission 'With maximum priority in the safety, to generate electricity by nuclear means with competitive quality and cost, sustained in our personnel's continuous overcoming and deep respect to the environment' and respecting our values (safety, responsibility by results, professional integrity, continuous improvement, team work, excellence in the performance, quality of service, protection to the environment its thought about our strategic objectives of the power station being born by this way one of them that it is the improvement program 'reduction of reloading times' looking for to be improves every day comparing us with the best plants of the world effectiveness all the processes in the power station that allowed us to measure our performance with the same parameters that settle down at international level as its are nuclear safety, industrial safety, radiological safety, capacity factor, readiness factor, cleaning of the power station attachment to procedures, attention to the detail and certainly to be competitive in the economic aspect. After analyzing the performance record of the power station, evaluating our technical, economic capacity, the location of the installation besides revising the international experiences was defined that one of the concepts that impact considerably so much to the capacity factors and readiness besides the dose and production cost is the duration of the reloading periods, for this reason work strategies were elaborated to be able to reach our goals of reloading days in less than 30 days, here are formed the carried out actions that they made us complete the four last reloading in less than 30 days. (Author)

  11. Three stops of fuel reloading with length of less 30 days in the Laguna Verde Central; Tres paradas de recarga de combustible con duracion de menos de 30 dias en la Central Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Lozano L, A. [Comision Federal de Electricidad, Central Laguna Verde, Subgerencia General de Operacion Planeacion, Veracruz (Mexico)]. e-mail: agustin.lozano@cfe.gob.mx

    2007-07-01

    The Laguna Verde Central having established as mission 'With maximum priority in the safety, to generate electricity by nuclear means with quality and cost competitive, sustained in our personnel's continuous overcoming and deep respect to the environment' and respecting our values (safety, responsibility by results, professional integrity, continuous improving, team working, excellence in the acting, quality of service, protection to the environment) they thought about our strategic objectives of the power station being born this way one of them that it is the program of improvement 'Reduction of reload times' looking for to be improves every day comparing us with the best plants in the world efficient all the processes in the power station that allowed us to measure our acting with the same parameters that settle down at international level like they are nuclear safety, industrial safety, radiological safety, capacity factor, readiness factor, cleaning of the power station attachment to procedures, attention to the detail and certainly to be competitive in the economic aspect. After analyzing the acting record of the power station, evaluating our technical capacity, economic, the location of the installation besides revising the international experiences it was defined that one of the concepts that impact considerably so much to the capacity and readiness factors besides the dose and production cost is the duration of the reload periods, for this reason they were elaborated work strategies to be able to reach our goal of reload days considered in being able to carry out them in less than 30 days, here the actions carried out that they made us complete the three last reloads in less than 30 days are captured. (Author)

  12. Interactive color graphics system for BWR fuel management

    International Nuclear Information System (INIS)

    Reese, A.P.

    1986-01-01

    An interactive color graphics system has been developed by the General Electric Company for fuel management engineers. The system consists of a Hewlett-Packard color graphics workstation in communication with a host mainframe. The system aids in such tasks as fuel cycle optimization, refueling bundle shuffle and control blade sequence design. Since being installed in 1983 turn-around time for a typical cycle reload and control blade pattern design has been reduced by a factor of four

  13. 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

  14. Additional guideline for the design of BWR reload patterns

    International Nuclear Information System (INIS)

    Burte, D.P.

    1990-01-01

    The problems of the high magnitude of the reactivity worths of control rods/notches and the thermal shocks to fuel due to control rod withdrawals are considered in this paper. These problems are shown to be mitigated if in addition to complying with the traditional guidelines for designing fuel bundles and reload patterns the proposed additional guideline (in the form of the condition that the shape of the K∞ distribution remains constant throughout the cycle) is also complied with. It is argued that application of the additional guideline for operating conditions ensures that the core can operate with a constant power profile throughout the cycle without control rod movement. Fixed (integral or stand-alone) burnable poisons are thought to be useful in this regard. Some comments on how this may be attempted are included. Additional advantages resulting from the additional guideline are discussed. (author)

  15. AUTOLOAD, an automatic optimal pressurized water reactor reload design system with an expert module

    International Nuclear Information System (INIS)

    Li, Z.; Levine, S.H.

    1994-01-01

    An automatic optimal pressurized water reactor (PWR) reload design expert system AUTOLOAD has been developed. It employs two important new techniques. The first is a new loading priority scheme that defines the optimal placement of the fuel in the core that has the maximum end-of-cycle state k eff . The second is a new power-shape-driven progressive iteration method for automatically determining the burnable poison (BP) loading in the fresh fuel assemblies. The Haling power distribution is used in converting the theoretically optimal solution into the practical design, which meets the design constraints for the given fuel assemblies. AUTOLOAD is a combination of C and FORTRAN languages. It requires only the required cycle length, the maximum peak normalized power, the BP type, the number of fresh fuel assemblies, the assembly burnup, and BP histories of the available fuel assemblies as its input. Knowledge-based modules have been built into the expert system computer code to perform all of the tasks involved in reloading a PWR. AUTOLOAD takes only ∼ 30 CPU min on an IBM 3090 600s mainframe to accomplish a practical reload design. A maximum of 12.5% fresh fuel enrichment saving is observed compared with the core used by the utility

  16. Reloading pattern optimization of VVER-1000 reactors in transient cycles using genetic algorithm

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • The genetic algorithm (GA) and the innovative weighting factors method were used. • The coupling of WIMSD5-B and CITATION-LDI2 neutronic codes with the thermohydraulic WERL code was employed. • Optimization of reloading patterns was carried out in two states. • First an arrangement with satisfactory excess reactivity and the flattest power distribution was searched. • Second, it is tried to obtain an arrangement with satisfactory safety threshold and the maximum K_e_f_f. - Abstract: The present paper proposes application of the genetic algorithm (GA) and the innovative weighting factor method to optimize the reloading pattern of Bushehr VVER-1000 reactor in the second cycle. To estimate the composition of fuel assemblies remaining from the first cycle and precisely calculate the objective parameters of each reloading pattern in the second cycle, coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic section was employed. Optimization of the reloading patterns was carried out in two states. To meet the mentioned objective, with application of the weighting factor method in the first state, the type and quantity of the loadable fresh assemblies were determined to enable the reactor core to maintain the core criticality over the entire cycle length. Afterwards, the genetic algorithm was used to optimize the reloading pattern of the reactor to obtain an arrangement with flat radial power distribution. In the second state, the optimization algorithm was free to select the type and number of fresh fuel assemblies to be able to search for an arrangement with the maximum effective multiplication factor and the safe power peaking factor. In addition, in order to ensure the safety and desirability of the proposed patterns in both states, a time-dependent examination of the thermo-neutronic behavior of the reactor core was carried out during the second cycle. With consideration of the new

  17. 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)

  18. 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)

  19. A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Varuttamaseni, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2017-09-30

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  20. A reload and startup plan for conversion of the NIST research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  1. Study of low leakage reload schedulle without burnable posion for Angra-1

    International Nuclear Information System (INIS)

    Sakai, M.; Dias, A.

    1989-01-01

    At the moment, there is a world trend to design larger cycles for PWR. Then the reload batches are increased, the enrichment in 235 U is increased and/or advanced fuel management strategies with radial low neutron leakage are applied. For the low leakage reloads of Angra-1 calculations were performed for different number of fuel assemblies for reaload batch, 32,36,40,44 and 48, from the 4th cycle up to equilibrium cycle for two different enrichments 3,4 W/O and 3,9 W/O in 235 U. The results showed that for the enrichments used without burnable posion it is possible to reach an increase in cycle lenghts between 3% and 8% for the same conditions. (author) [pt

  2. Data bank usage in reload design and licensing

    International Nuclear Information System (INIS)

    Goudey, J.L.; Hansen, E.C.; Scigliano, S.M.; Williams, R.D.

    1986-01-01

    In 1977 the Nuclear Energy Business Operations of General Electric Company (GE) began a major project to automate sequential execution of the data transfer between the various computer programs used in performing calculations to support design, release, licensing, and core management of fuel used in boiling water reactors (BWRs). A centralized and controlled data bank was designed and implemented to complement the data management system and to achieve the following objectives: (1) enhance the quality and reliability of engineering data used for design and licensing of BWR fuel; (2) provide for traceability and long-term retrievability of engineering data as required by 10CFR50, Appendix B; (3) standardize the location and minimize the redundancy of engineering data; and (4) make engineering data readily available to all individuals and computer programs with a need for it. The structure of this data bank, which has become known as the BWR Engineering Data Bank or BWR/EDB, was purposefully left flexible and expandable with the ability to accommodate numerical, logical, and textual data. The BWR/EDB has been used by GE during fuel release, fuel and core design, reload licensing, and core management activities for 30 to 40 commercial power reactors over the past several years

  3. 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)

  4. 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)

  5. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Egan, M.R.

    1984-01-01

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behaviour and physical requirements of operating cycle sequences and fueling strategies having practical use in fuel management. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and manoeuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy. Numerical evaluations of degenerate equilibrium cycle sequences are then performed for a typical PWR core, and accompanying fuel cycle costs are calculated. The impact of design and operational limits as constraints on the performance mappings for this reactor are also studied with respect to achieving improved cost performance from the once-through fuel cycle. The dynamics of transition cycle sequences are then examined using the generalized theory. Proof of the existence of non-degenerate equilibrium cycle sequences is presented when the mechanics of the fixed reload batch size strategy are developed analytically for transition sequences. Finally, an analysis of the fixed reload enrichment strategy demonstrates the potential for convergence of the transition sequence to a fully degenerate equilibrium sequence. (author)

  6. Plutonium assemblies in reload 1 of the Dodewaard Reactor

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.; Leenders, L.; Mostert, P.

    1977-01-01

    Since 1963, Belgonucleaire has been developing the design of plutonium assemblies of the island type (i.e., plutonium rods inserted in the control zone of the assembly and enriched uranium rods at the periphery) for light water reactors. The application to boiling water reactors (BWRs) led to the introduction, in April 1971, of two prototype plutonium island assemblies in the Dodewaard BWR (The Netherlands): Those assemblies incorporating plutonium in 42 percent of the rods are interchangeable with standard uranium assemblies of the same reload. Their design, which had to meet these criteria, was performed using the routine order in use at Belgonucleaire; experimental checks included a mock-up configuration simulated in the VENUS critical facility at Mol and open-vessel cold critical experiments performed in the Dodewaard core. The pelleted plutonium rods were fabricated and controlled by Belgonucleaire following the manufacturing procedures developed at the production plant. In one of the assemblies, three vibrated plutonium fuel rods with a lower fuel density were introduced in the three most highly rated positions to reduce the power rating. Those plutonium assemblies experienced peak pellet ratings up to 535 W/cm and were discharged in April 1974 after having reached a mean burnup of approximately 21,000 MWd/MT. In-core instrumentation during operation, visual examinations, and reactivity substitution experiments during reactor shutdown did not indicate any special feature for those assemblies compared to the standard uranium assemblies, thereby demonstrating their interchangeability

  7. Kinetics and Muscle Activity Patterns during Unweighting and Reloading Transition Phases in Running.

    Directory of Open Access Journals (Sweden)

    Patrick Sainton

    Full Text Available Amongst reduced gravity simulators, the lower body positive pressure (LBPP treadmill is emerging as an innovative tool for both rehabilitation and fundamental research purposes as it allows running while experiencing reduced vertical ground reaction forces. The appropriate use of such a treadmill requires an improved understanding of the associated neuromechanical changes. This study concentrates on the runner's adjustments to LBPP-induced unweighting and reloading during running. Nine healthy males performed two running series of nine minutes at natural speed. Each series comprised three sequences of three minutes at: 100% bodyweight (BW, 60 or 80% BW, and 100% BW. The progressive unweighting and reloading transitions lasted 10 to 15 s. The LBPP-induced unweighting level, vertical ground reaction force and center of mass accelerations were analyzed together with surface electromyographic activity from 6 major lower limb muscles. The analyses of stride-to-stride adjustments during each transition established highly linear relationships between the LBPP-induced progressive changes of BW and most mechanical parameters. However, the impact peak force and the loading rate systematically presented an initial 10% increase with unweighting which could result from a passive mechanism of leg retraction. Another major insight lies in the distinct neural adjustments found amongst the recorded lower-limb muscles during the pre- and post-contact phases. The preactivation phase was characterized by an overall EMG stability, the braking phase by decreased quadriceps and soleus muscle activities, and the push-off phase by decreased activities of the shank muscles. These neural changes were mirrored during reloading. These neural adjustments can be attributed in part to the lack of visual cues on the foot touchdown. These findings highlight both the rapidity and the complexity of the neuromechanical changes associated with LBPP-induced unweighting and reloading

  8. Kinetics and Muscle Activity Patterns during Unweighting and Reloading Transition Phases in Running

    Science.gov (United States)

    Sainton, Patrick; Nicol, Caroline; Cabri, Jan; Barthèlemy-Montfort, Joëlle; Chavet, Pascale

    2016-01-01

    Amongst reduced gravity simulators, the lower body positive pressure (LBPP) treadmill is emerging as an innovative tool for both rehabilitation and fundamental research purposes as it allows running while experiencing reduced vertical ground reaction forces. The appropriate use of such a treadmill requires an improved understanding of the associated neuromechanical changes. This study concentrates on the runner’s adjustments to LBPP-induced unweighting and reloading during running. Nine healthy males performed two running series of nine minutes at natural speed. Each series comprised three sequences of three minutes at: 100% bodyweight (BW), 60 or 80% BW, and 100% BW. The progressive unweighting and reloading transitions lasted 10 to 15 s. The LBPP-induced unweighting level, vertical ground reaction force and center of mass accelerations were analyzed together with surface electromyographic activity from 6 major lower limb muscles. The analyses of stride-to-stride adjustments during each transition established highly linear relationships between the LBPP-induced progressive changes of BW and most mechanical parameters. However, the impact peak force and the loading rate systematically presented an initial 10% increase with unweighting which could result from a passive mechanism of leg retraction. Another major insight lies in the distinct neural adjustments found amongst the recorded lower-limb muscles during the pre- and post-contact phases. The preactivation phase was characterized by an overall EMG stability, the braking phase by decreased quadriceps and soleus muscle activities, and the push-off phase by decreased activities of the shank muscles. These neural changes were mirrored during reloading. These neural adjustments can be attributed in part to the lack of visual cues on the foot touchdown. These findings highlight both the rapidity and the complexity of the neuromechanical changes associated with LBPP-induced unweighting and reloading during running

  9. A reverse depletion method for pressurized water reactor core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.; Kin, Y.J.

    1986-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. The large numbers of burnable poison pins used to control the power peaking at the in-board fresh fuel positions have introduced an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of burnable poison pins in the fresh fuel. A new method for performing core design more suitable for low-leakage fuel management is reported. A procedure was developed that uses the wellknown ''Haling depletion'' to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and burnable poison distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provided for the maximum utility to PWR core reload design

  10. Spent fuel bundle counter sequence error manual - RAPPS (200 MW) NGS

    International Nuclear Information System (INIS)

    Nicholson, L.E.

    1992-01-01

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message typically contains adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  11. Spent fuel bundle counter sequence error manual - KANUPP (125 MW) NGS

    International Nuclear Information System (INIS)

    Nicholson, L.E.

    1992-01-01

    The Spent Fuel Bundle Counter (SFBC) is used to count the number and type of spent fuel transfers that occur into or out of controlled areas at CANDU reactor sites. However if the transfers are executed in a non-standard manner or the SFBC is malfunctioning, the transfers are recorded as sequence errors. Each sequence error message may contain adequate information to determine the cause of the message. This manual provides a guide to interpret the various sequence error messages that can occur and suggests probable cause or causes of the sequence errors. Each likely sequence error is presented on a 'card' in Appendix A. Note that it would be impractical to generate a sequence error card file with entries for all possible combinations of faults. Therefore the card file contains sequences with only one fault at a time. Some exceptions have been included however where experience has indicated that several faults can occur simultaneously

  12. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  13. Reload Startup Physics Tests for Tianwan Nuclear Power station

    International Nuclear Information System (INIS)

    Yang Xiaoqiang; Li Wenshuang; Li Youyi; Yao Jinguo; Li Zaipeng Jiangsu

    2010-01-01

    This paper briefly describes the test purposes, test items, test schedules and test equipment's for reload startup physics test's on Unit 1 and 2 of Tianwan Nuclear Power station. Then, an overview of the previous thrice tests and evaluations on the tests results are presented. In the end, the paper shows the development and work direction of optimization project for reload startup physics tests on Unit 1 and 2 of Tianwan Nuclear Power station. (Authors)

  14. The scale analysis sequence for LWR fuel depletion

    International Nuclear Information System (INIS)

    Hermann, O.W.; Parks, C.V.

    1991-01-01

    The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

  15. Study of different fitness functions with safety restriction for nuclear reactor reload problem using QDPSO

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo C. de, E-mail: paulocaixeta@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear; Lima, Alan M.M. de; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Nuclear Reactor Reload Problem (NRRP) is a classical problem in Nuclear Engineering that has been studied for more than 40 years, which focuses on the economics and safety of the Nuclear Power Plant (NPP). This problem consists in searching for the best loading pattern of fuel assemblies (FA) in the core, aiming to determine the permutation of fuel assemblies that optimizes the uranium utilization, with fitness function evaluated according to specific criteria and methods of nuclear reactor physics, such as the maximum mean power peak and the boron concentration. In this article will be presented different methodologies to obtain a representative fitness function for NRRP, where Quantum particle Swarm optimization (QPSO) was used to determine which one gives the best array of fuel assemblies that will make the maximum EFPD (Effective Full Power Days) with the least computational effort. In this approach, as well as others in literature, was not used Burnable Poison in the simulations and the results will be compared in relation of the maximization of the cycle length considering the boron concentration yield by the reactor physics code, to make sure that the configuration is valid from a safety point of view. This paper was based on Angra 1's seventh reload cycle. (author)

  16. 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.

  17. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors; Actualizacion de los costos de combustible nuclear de la recarga de equilibrio de los reactores ABWR y EPR

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, R.F. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: rortega@fi-b.unam.mx

    2008-07-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U{sub 3O}8 in January, 2005 to a maximum of US$137.00 dollars by Ib U{sub 3}O{sub 8} by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U{sub 3}O{sub 8} such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  18. Cuckoo Search with flight of Levy applied to the problem of reload of fuels in nuclear reactors; Busca Cuco com voo de Levy aplicado ao problema de recarga de combustiveis em reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Patrick V.; Nast, Fernando N.; Schirru, Roberto; Meneses, Anderson A.M., E-mail: patrickvs@hotmail.com, E-mail: fernandonnast@gmail.com, E-mail: anderson.meneses@pq.cnpq.br, E-mail: schirru@lmp.ufrj.br [Universidade Federal do Oeste do Para (UFOPA), Santarem, PA (Brazil). Instituto de Engenharia e Geociencias. Programa de Ciencia e Tecnologia; Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ) RJ (Brazil). Programa de Engenharia Nuclear

    2017-11-01

    Intra-Nuclear Fuel Management Optimization is a complex combinatorial problem of the NP-difficult type, associated with the refueling process of a nuclear reactor, which aims to extend the cycle of operation by determining loading patterns, obeying safety margins. In addition to the combinatorial problem, we have the aspect of calculations of reactor physics, which increases the difficult of OGCIN. Methods that are proving effective when applied to OGCIN are the algorithms belonging to the swarm intelligence paradigm. A new member of this paradigm is Cuckoo Search (CS), which has shown results promising when applied to optimization issues. The CS is based on the litter parasitism of some cuckoo species combined with the Levy flight behavior of some birds. In the present work we present the results of the application of CS to OGCIN, and compare them to the results obtained by the application of ABC.

  19. An approach using quantum ant colony optimization applied to the problem of nuclear reactors reload

    International Nuclear Information System (INIS)

    Silva, Marcio H.; Lima, Alan M.M. de; Schirru, Roberto; Medeiros, J.A.C.C.

    2009-01-01

    The basic concept behind the nuclear reactor fuel reloading problem is to find a configuration of new and used fuel elements, to keep the plant working at full power by the largest possible duration, within the safety restrictions. The main restriction is the power peaking factor, which is the limit value for the preservation of the fuel assembly. The QACO A lfa algorithm is a modified version of Quantum Ant Colony Optimization (QACO) proposed by Wang et al, which uses a new actualization method and a pseudo evaporation step. We examined the QACO A lfa behavior associated to physics of reactors code RECNOD when applied to this problem. Although the QACO have been developed for continuous functions, the binary model used in this work allows applying it to discrete problems, such as the mentioned above. (author)

  20. Degradation resistant fuel cladding materials and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Montes, J. [ENUSA, Madrid (Spain)

    1995-12-31

    GE has been producing the degradation resistant cladding (zirconium liner and zircaloy-2 surface larger) described here with the cooperation of its primary zirconium vendors since the beginning of 1994. Approximately 24 fuel reloads, or in excess of 250,000 fuel rods, have been produced using this material by GE. GE has also produced tubing for one reload of fuel that is currently being produced by its technology affiliate ENUSA. (orig./HP)

  1. 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)

  2. Validity of single-cycle objective functions for multicycle reload design optimization

    International Nuclear Information System (INIS)

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

    1993-01-01

    Beyond the equilibrium cycle scoping calculations used for determining numbers of feed assemblies and enrichment estimates, multicycle reload design currently consists of stagewise optimization of single-cycle core loading patterns, typically extending over a short-term planning horizon of perhaps three reload cycles. Particularly in transition cycles, however, optimizing a loading pattern over a single cycle for a stated objective, such as minimum core leakage, may have an adverse impact on subsequent cycles. The penalties paid may be in the form of reduced thermal margin or an increase in feed enrichment due to insufficient reactivity carryover from the open-quotes optimizedclose quotes cycle. In view of current practices, a study was performed that examined the behavior of the loading pattern as a function of the objective functions selected as implemented in the stagewise optimization of single-cycle core loading patterns from initial transition cycle through equilibrium using the FORMOSA-P code. The objective functions studied were region average discharge burnup maximization (with enrichment search) and feed enrichment minimization. It is noted at the beginning that the maximization of region average discharge has no meaning for the equilibrium cycle because region average discharge burnup is explicitly set by the feed size and cycle length independent of the loading pattern. In the nonequilibrium cycle, however, it was reasoned that this objective would provide the maximum reactivity carryover throughout the transition and thus have a direct effect on minimizing the multicycle levelized fuel cost

  3. AC-600 reactor reloading pattern optimization by using genetic algorithms

    International Nuclear Information System (INIS)

    Wu Hongchun; Xie Zhongsheng; Yao Dong; Li Dongsheng; Zhang Zongyao

    2000-01-01

    The use of genetic algorithms to optimize reloading pattern of the nuclear power plant reactor is proposed. And a new encoding and translating method is given. Optimization results of minimizing core power peak and maximizing cycle length for both low-leakage and out-in loading pattern of AC-600 reactor are obtained

  4. Reload of cobalt 60 for the J S-6500 irradiator

    International Nuclear Information System (INIS)

    Torres C, G.; Mayoral G, V.M.

    1991-01-01

    The present work has the purpose to describe the activities of the reloads program of the industrial irradiator J S-6500, elaborated for satisfying part of the demand of services and as a first step, to guide the decision making by the part of the ININ authorities in front of a wide market of this service. (Author)

  5. 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

  6. Automatic multi-cycle reload design of pressurized water reactor using particle swarm optimization algorithm and local search

    International Nuclear Information System (INIS)

    Lin, Chaung; Hung, Shao-Chun

    2013-01-01

    Highlights: • An automatic multi-cycle core reload design tool, which searches the fresh fuel assembly composition, is developed. • The search method adopts particle swarm optimization and local search. • The design objectives are to achieve required cycle energy, minimum fuel cost, and the satisfactory constraints. • The constraints include the hot zero power moderator temperature coefficient and the hot channel factor. - Abstract: An automatic multi-cycle core reload design tool, which searches the fresh fuel assembly composition, is developed using particle swarm optimization and local search. The local search uses heuristic rules to change the current search result a little so that the result can be improved. The composition of the fresh fuel assemblies should provide the required cycle energy and satisfy the constraints, such as the hot zero power moderator temperature coefficient and the hot channel factor. Instead of designing loading pattern for each FA composition during search process, two fixed loading patterns are used to calculate the core status and the better fitness function value is used in the search process. The fitness function contains terms which reflect the design objectives such as cycle energy, constraints, and fuel cost. The results show that the developed tool can achieve the desire objective

  7. Review of modifications performed to core monitoring systems related to core reloading

    International Nuclear Information System (INIS)

    Carew, J.F.; Diamond, D.J.

    1978-09-01

    The recent increase in the number of licensees selecting new fuel suppliers for reload cycles has resulted in a trend toward Core Monitoring Systems (CMS) for which the cycle dependent data and the CMS software are supplied by different vendors. At the request of and under the direction of the Division of Operating Reactors, USNRC, a review of the qualification and documentation for these CMS has been made. Several potential problem areas in the determination of CMS cycle dependent input involving empirical normalizations and relatively complex neutronic analysis were identified. As representative of present qualification and documentation practices, Yankee Atomic Electric Co., Virginia Electric Power Co., Nuclear Associations International, Exxon Nuclear Co., Northeast Utilities Service Co. and Jersey Central Power and Light were selected and reviewed in detail

  8. 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

  9. Biodegradation of Jet Fuel-4 (JP-4) in Sequencing Batch Reactors

    Science.gov (United States)

    1992-06-01

    nalw~eo %CUMENTATION PAGE__ _ _ _ _ _ _ _ _O 74S Ab -A258 020 L AW POi~W6 DATI .~ TYP AIMqm ,-& 0 U. glbs A~ I ma"&LFUN Mu BIODEGRADATION OF JET FUEL...Specific Objectives of This Proposal Are: 1. To assess the ability of C. resinae , P. chrysosporium and selected bacterial consortia to degrade individual...chemical components of JP-4. 2. To develop a sequencing batch reactor that utilizes C. resinae to degrade chemical components of JP-4 in contaminated

  10. Loading pattern optimization with maximum utilization of discharging fuel employing adaptively constrained discontinuous penalty function

    International Nuclear Information System (INIS)

    Park, T. K.; Joo, H. G.; Kim, C. H.

    2010-01-01

    In order to find the most economical loading pattern (LP) considering multi-cycle fuel loading, multi-objective fuel LP optimization problems are examined by employing an adaptively constrained discontinuous penalty function (ACDPF) method. This is an improved method to simplify the complicated acceptance logic of the original DPF method in that the stochastic effects caused by the different random number sequence can be reduced. The effectiveness of the multi-objective simulated annealing (SA) algorithm employing ACDPF is examined for the reload core LP of Cycle 4 of Yonggwang Nuclear Unit 4. Several optimization runs are performed with different numbers of objectives consisting of cycle length and average burnup of fuels to be discharged or reloaded. The candidate LPs obtained from the multi-objective optimization runs turn out to be better than the reference LP in the aspects of cycle length and utilization of given fuels. It is note that the proposed ACDPF based MOSA algorithm can be a practical method to obtain an economical LP considering multi-cycle fuel loading. (authors)

  11. 24-month fuel cycles

    International Nuclear Information System (INIS)

    Rosenstein, R.G.; Sipes, D.E.; Beall, R.H.; Donovan, E.J.

    1986-01-01

    Twenty-four month reload cycles can potentially lessen total power generation costs. While 24-month cores increase purchased fuel costs, the longer cycles reduce the number of refueling outages and thus enhance plant availability; men-rem exposure to site personnel and other costs associated with reload core design and licensing are also reduced. At dual unit sites an operational advantage can be realized by refueling each plant alternately on a 1-year offset basis. This results in a single outage per site per year which can be scheduled for off-peak periods or when replacement power costs are low

  12. Parameterless evolutionary algorithm applied to the nuclear reload problem

    International Nuclear Information System (INIS)

    Caldas, Gustavo Henrique Flores; Schirru, Roberto

    2008-01-01

    In this work, an evolutionary algorithm with no parameters called FPBIL (parameter free PBIL) is developed based on PBIL (population-based incremental learning). Moreover, the analysis reveals how the parameters from PBIL can be replaced by self-adaptable mechanisms which appear from the radically different form by which the evolution is processed. Despite the advantages, the FPBIL reveals itself compact and relatively modest in the use of computational resources. The FPBIL is then applied to the nuclear reload problem. The experimental results observed are compared to those of other works and corroborate to affirm the superiority of the new algorithm

  13. A comparison between genetic algorithms and neural networks for optimizing fuel recharges in BWR

    International Nuclear Information System (INIS)

    Ortiz J, J.; Requena, I.

    2002-01-01

    In this work the results of a genetic algorithm (AG) and a neural recurrent multi state network (RNRME) for optimizing the fuel reload of 5 cycles of the Laguna Verde nuclear power plant (CNLV) are presented. The fuel reload obtained by both methods are compared and it was observed that the RNRME creates better fuel distributions that the AG. Moreover a comparison of the utility for using one or another one techniques is make. (Author)

  14. Clopidogrel reloading for patients with acute myocardial infarction already on clopidogrel therapy.

    Science.gov (United States)

    Doll, Jacob A; Li, Shuang; Chiswell, Karen; Roe, Matthew T; Kosiborod, Mikhail; Scirica, Benjamin M; Wang, Tracy Y

    2018-01-14

    We sought to determine the association of clopidogrel reloading with in-hospital bleeding and mortality in contemporary practice. We examined clopidogrel reloading for ST-segment elevation myocardial infarction (STEMI) and non-ST-segment elevation myocardial infarction (NSTEMI) patients on pre-admission clopidogrel therapy in the ACTION Registry-GWTG from 2009 to 2014. We used inverse probability weighted propensity adjustment to compare in-hospital mortality and major bleeding risks between patients reloaded in the first 24 h with ≥300 mg of clopidogrel vs. those continued on a maintenance (therapy who were admitted with STEMI, 9369 (75.8%) received a loading dose. Of 39 158 patients with NSTEMI, 10 144 (25.9%) were reloaded. Reloaded patients were younger, had fewer comorbid conditions, and were more likely to be treated with primary PCI (STEMI) or an early invasive strategy (NSTEMI). Risks of major bleeding were not significantly different between patients with and without reloading, whether presenting with STEMI (OR 0.98, 95% CI 0.85-1.13) or NSTEMI (OR 1.00, 95% CI 0.90-1.11). Among STEMI patients, clopidogrel reloading was associated with lower risks of in-hospital mortality (OR 0.80, 95% CI 0.66-0.96), however no significant mortality difference was observed among NSTEMI patients (OR 1.13, 95% CI 0.93-1.37). Clopidogrel reloading occurs frequently among MI patients who are on pre-admission clopidogrel therapy, particularly among STEMI patients. We did not observe increased bleeding or mortality risk with clopidogrel reloading, and therefore reloading could be safe for most MI patients. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2017. For permissions, please email: journals.permissions@oup.com.

  15. New strategies of reloads design and models of control bars in boiling water reactors; Nuevas estrategias de diseno de recargas y de patrones de barras de control en reactores de agua en ebullicion

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Perusquia del Cueto, R., E-mail: alejandro.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2011-11-15

    In this work the results obtained when analyzing new strategies in the reload designs of nuclear fuel and models of control bars, for boiling water reactors are presented. The idea is to analyze the behaviour of the reactor during an operation cycle, when the heuristic rules are not used (commonly used by expert engineers in both designs). Specifically was analyzed the rule of low leak and the load strategy Control Cell Core for the design of a fuel reload. In a same way was analyzed the rule of prohibiting the use of the intermediate positions in the control bars, as well as the construction of bar models based on load strategies type Control Cell Core. In the first analysis a balance and transition cycle were used. For the second analysis only a transition cycle was used, firstly with the reloads designed in the first analysis and later on with reloads built by other methods. For the simulation of the different configurations proposed in both cases, was used the code Simulate-3. To obtain the designs in both studies, the heuristic techniques or neural networks and taboo search were used. The obtained results show that it can be omitted of some rules used in the ambit for the mentioned designs and even so to obtain good results. To carry out this investigation was used Dell work station under Li nux platform. (Author)

  16. New techniques for designing the initial and reload cores with constant long cycle lengths

    International Nuclear Information System (INIS)

    Shi, Jun; Levine, Samuel; Ivanov, Kostadin

    2017-01-01

    Highlights: • New techniques for designing the initial and reload cores with constant long cycle lengths are developed. • Core loading pattern (LP) calculations and comparisons have been made on two different designs. • Results show that significant savings in fuel costs can be accrued if a non-low leakage LP design strategy is enacted. - Abstract: Several utilities have increased the output power of their nuclear power plant to increase their income and profit. Thus, the utility increases the power density of the reactor, which has other consequences. One consequence is to increase the depletion of the fuel assemblies (FAs) and reduce the end-of-cycle (EOC) sum of fissionable nuclides in each FA, ∑_E_O_C. The power density and the ∑_E_O_C remaining in the FAs at EOC must be sufficiently large in many FAs when designing the loading pattern, LP, for the first and reload cycles to maintain constant cycle lengths at minimum fuel cost. Also of importance is the cycle length as well as several other factors. In fact, the most important result of this study is to understand that the ∑_E_O_Cs in the FAs must be such that in the next cycle they can sustain the energy during depletion to prevent too much power shifting to the fresh FAs and, thus, sending the maximum peak pin power, PPP_m_a_x, above its constraint. This paper presents new methods for designing the LPs for the initial and follow on cycles to minimize the fuel costs. Studsvik’s CMS code system provides a 1000 MWe LP design in their sample inputs, which is applied in this study. The first 3 cycles of this core are analyzed to minimize fuel costs, and all three cycles have the same cycle length of ∼650 days. Cycle 1 is designed to allow many used FAs to be loaded into cycles 2 and 3 to reduce their fuel costs. This could not be achieved if cycle 1 was a low leakage LP (Shi et al., 2015). Significant fuel cost savings are achieved when the new designs are applied to the higher leakage LP designs

  17. A new evolutionary algorithm with LVQ learning for the optimization of combinatory problems as a reload of nuclear reactors

    International Nuclear Information System (INIS)

    Machado, Marcelo Dornellas

    1999-04-01

    Genetic algorithms are biologically motivated adaptive systems which have been used, with good results, for function optimization. In this work, a new learning mode, to be used by the Population-Based Incremental Learning (PBIL) algorithm, who combines mechanisms of standard genetic algorithm with simple competitive learning, has the aim to build a new evolutionary algorithm to be used in optimization of numerical problems and combinatorial problems. This new learning mode uses a variable learning rate during the optimization process, constituting a process know as proportional reward. The development of this new algorithm aims its application in the optimization of reload problem of PWR nuclear reactors. This problem can be interpreted as search of a load pattern to be used in the nucleus of the reactor in order to increase the useful life of the nuclear fuel. For the test, two classes of problems are used: numerical problems and combinatorial problem, the major interest relies on the last class. The results achieved with the tests indicate the applicability of the new learning mode, showing its potential as a developing tool in the solution of reload problem. (author)

  18. Levitation force relaxation under reloading in a HTS Maglev system

    International Nuclear Information System (INIS)

    He Qingyong; Wang Jiasu; Wang Suyu; Wang Jiansi; Dong Hao; Wang Yuxin; Shao Senhao

    2009-01-01

    The loading capacity of the high-temperature superconducting (HTS) Maglev vehicle is an important parameter in the practical application. It is closely related to the levitation force of the HTS bulk. Many papers reported that the levitation force showed the relaxation characteristic. Because different loads cause different levitation gaps and different applied magnetic fields, the levitation force relaxations under the different loads are not the same. In terms of cylindrical YBCO bulk levitated over the permanent magnetic guideway, the relationship between the levitation force relaxation and the reloading is investigated experimentally in this paper. The decrement, the decrement rate and the relaxation rate of the levitation force are calculated, respectively. This work might be helpful for studying the loading capacity of the HTS Maglev vehicle

  19. Levitation force relaxation under reloading in a HTS Maglev system

    Energy Technology Data Exchange (ETDEWEB)

    He Qingyong [Applied Superconductivity Laboratory, M/S 152, Southwest Jiaotong University, Chengdu, Sichuan 610031 (China)], E-mail: hedoubling@gmail.com; Wang Jiasu; Wang Suyu; Wang Jiansi; Dong Hao; Wang Yuxin; Shao Senhao [Applied Superconductivity Laboratory, M/S 152, Southwest Jiaotong University, Chengdu, Sichuan 610031 (China)

    2009-02-01

    The loading capacity of the high-temperature superconducting (HTS) Maglev vehicle is an important parameter in the practical application. It is closely related to the levitation force of the HTS bulk. Many papers reported that the levitation force showed the relaxation characteristic. Because different loads cause different levitation gaps and different applied magnetic fields, the levitation force relaxations under the different loads are not the same. In terms of cylindrical YBCO bulk levitated over the permanent magnetic guideway, the relationship between the levitation force relaxation and the reloading is investigated experimentally in this paper. The decrement, the decrement rate and the relaxation rate of the levitation force are calculated, respectively. This work might be helpful for studying the loading capacity of the HTS Maglev vehicle.

  20. Nuclear fuel cycle activities with an utility

    International Nuclear Information System (INIS)

    Schwarz, E.

    1977-01-01

    The lecture will deal with the following topics: Fuel requirements: establishing fuel requirements - first core - reloads. Calculation of required uranium and separation work: reload planning - long term - short term - during refuelling; exactness of calculations: contracts: 1) Uranium and conversion; 2) Enrichment services; 3) Fuel elements; 4) Ownership; 5) Accidential loss of material; 6) Flexibility in time and amounts; 7) Specifications, surcharges; 8) Terms of payment; 9) Fuel containers, ownership, retransport; fuel reserves: 1) Natural uranium (concentrates or reserves in the ground); 2) Enriched uranium; 3) Fuel elements; 4) Cost of reserves; 5) Exchange in case of need. Handling of contracts: 1) Schedule for deliveries; Notes for deliveries; 3) Fuel accounting and balance; 4) Formalities (export and import licenses, customs etc.). Fuel cost: 1) Prices; 2) Fuel cost calculations for comparison of bids and cost forecast. (orig.) [de

  1. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali Santos de; Schirru, Roberto, E-mail: ioliveira@con.ufrj.b, E-mail: schirru@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  2. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    International Nuclear Information System (INIS)

    Oliveira, Iona Maghali Santos de; Schirru, Roberto

    2011-01-01

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  3. 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)

  4. 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)

  5. SEDRIO/INCORE, an automatic optimal loading pattern search system for PWR NPP reload core using an expert system

    International Nuclear Information System (INIS)

    Xian Chunyu; Zhang Zongyao

    2003-01-01

    The expert knowledge library for Daya Bay and Qinshan phase II NPP has been established based on expert knowledge, and the reload core loading pattern heuristic search is performed. The in-core fuel management code system INCORE that has been used in engineering design is employed for neutron calculation, and loading pattern is evaluated by using of cycle length and core radial power peaking factor. The developed system SEDRIO/INCORE has been applied in cycle 4 for unit 2 of Daya Bay NPP and cycle 4 for Phase II in Qinshan NPP. The application demonstrated that the loading patterns obtained by SEDRIO/INCORE system are much better than reference ones from the view of the radial power peak and the cycle length

  6. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  7. Tailoring Vantage 5 (fuel) to suit each operator's need

    Energy Technology Data Exchange (ETDEWEB)

    Chapin, D L; Secker, J R [Westinghouse Electric Corp., Philadelphia, PA (USA)

    1990-03-01

    By the end of 1989, Westinghouse Vantage 5 fuel had been reloaded into 36 nuclear power plants. The fuel offers a number of features operators can choose from to suit their own particular needs. Experience so far has shown the fuel to have performed well, with coolant activity levels remaining low. (author).

  8. Core-state models for fuel management of equilibrium and transition cycles in pressurized water reactors

    International Nuclear Information System (INIS)

    Aragones, J.M.; Martinez-Val, J.M.; Corella, M.R.

    1977-01-01

    Fuel management requires that mass, energy, and reactivity balance be satisfied in each reload cycle. Procedures for selection of alternatives, core-state models, and fuel cost calculations have been developed for both equilibrium and transition cycles. Effective cycle lengths and fuel cycle variables--namely, reload batch size, schedule of incore residence for the fuel, feed enrichments, energy sharing cycle by cycle, and discharge burnup and isotopics--are the variables being considered for fuel management planning with a given energy generation plan, fuel design, recycling strategy, and financial assumptions

  9. Increasing TRIGA fuel lifetime with 12 wt.% U TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Naughton, W F; Cenko, M J; Levine, S H; Witzig, W F [Pennsylvania State University (United States)

    1974-07-01

    In-core fuel management studies have been performed for the Penn State Breazeale Reactor (PSBR) wherein 12 wt % U fuel elements are used to replace the standard 8.5 wt % U TRIGA fuel. The core configuration used to develop a calculational model was a 90-element hexagonal array, which is representative of the PSBR core, and consists of five hexagonal rings surrounding a central thimble containing water. The technique employed for refueling the core fully loaded with 8.5 wt % U fuel involves replacing 8.5 wt % U fuel with 12 wt % U fuel using an in-out reloading scheme. A batch reload consists of 6 new 12 wt % U fuel elements. Placing the 12 wt % U fuel in the B ring produces fuel temperatures ({approx}450 {sup o}C) that are well below the 800{sup o}C maximum limitation when the PSBR is operating at its maximum allowed power of 1 Megawatt. The advantages of using new 12 wt % U fuel to replace the burned up 8.5 wt % U fuel in the B ring over refueling strictly with 8.5 wt % U-Zr TRIGA fuel are clearly delineated in Table 1 where cost calculations used the General Atomic pre-1972 prices for TRIGA fuel, i.e., $1500 and $1650 for an 8.5 and 12 wt % U fuel element, respectively. Experimental results obtained to date utilizing the 12 wt % U fuel elements agree with the computed results. (author)

  10. Accelerated fuel depreciation as an economic incentive for low-leakage fuel management

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    An analysis is presented which evaluates the tax depreciation advantage which results from the increased rate of fuel depletion achieved in the current low-leakage fuel-management LWR core reload designs. An analytical fuel-cycle cost model is used to examine the important cost parameters which are then validated using the fuel-cycle cost code CINCAS and data from the Maine Yankee PWR. Results show that low-leakage fuel management, through the tax depreciation advantage from accelerated fuel depletion, provides an improvement of several percent in fuel-cycle costs compared to traditional out-in fuel management and a constant fuel depletion rate. (author)

  11. Operating experience with Exxon nuclear advanced fuel assembly and fuel cycle designs in PWRs

    International Nuclear Information System (INIS)

    Skogen, F.B.; Killgore, M.R.; Holm, J.S.; Brown, C.A.

    1986-01-01

    Exxon Nuclear Company (ENC) has achieved a high standard of performance in its supply of fuel reloads for both BWRs and PWRs, while introducing substantial innovations aimed at realization of improved fuel cycle costs. The ENC experience with advanced design features such as the bi-metallic spacer, the dismountable upper tie plate, natural uranium axial blankets, optimized water-to-fuel designs, annular pellets, gadolinia burnable absorbers, and improved fuel management scenarios, is summarized

  12. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel

    International Nuclear Information System (INIS)

    Francois, J.L.; Nunez C, A.

    2003-01-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  13. BWR fuel cycle optimization using neural networks

    International Nuclear Information System (INIS)

    Ortiz-Servin, Juan Jose; Castillo, Jose Alejandro; Pelta, David Alejandro

    2011-01-01

    Highlights: → OCONN a new system to optimize all nuclear fuel management steps in a coupled way. → OCON is based on an artificial recurrent neural network to find the best combination of partial solutions to each fuel management step. → OCONN works with a fuel lattices' stock, a fuel reloads' stock and a control rod patterns' stock, previously obtained with different heuristic techniques. → Results show OCONN is able to find good combinations according the global objective function. - Abstract: In nuclear fuel management activities for BWRs, four combinatorial optimization problems are solved: fuel lattice design, axial fuel bundle design, fuel reload design and control rod patterns design. Traditionally, these problems have been solved in separated ways due to their complexity and the required computational resources. In the specialized literature there are some attempts to solve fuel reloads and control rod patterns design or fuel lattice and axial fuel bundle design in a coupled way. In this paper, the system OCONN to solve all of these problems in a coupled way is shown. This system is based on an artificial recurrent neural network to find the best combination of partial solutions to each problem, in order to maximize a global objective function. The new system works with a fuel lattices' stock, a fuel reloads' stock and a control rod patterns' stock, previously obtained with different heuristic techniques. The system was tested to design an equilibrium cycle with a cycle length of 18 months. Results show that the new system is able to find good combinations. Cycle length is reached and safety parameters are fulfilled.

  14. A comparison between genetic algorithms and neural networks for optimizing fuel recharges in BWR; Una comparacion entre algoritmos geneticos y redes neuronales para optimizar recargas de combustible en BWR's

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz J, J. [Instituto Nacional de Investigaciones Nucleares, Depto. Sistemas Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico); Requena, I. [Universidad de Granada (Spain)

    2002-07-01

    In this work the results of a genetic algorithm (AG) and a neural recurrent multi state network (RNRME) for optimizing the fuel reload of 5 cycles of the Laguna Verde nuclear power plant (CNLV) are presented. The fuel reload obtained by both methods are compared and it was observed that the RNRME creates better fuel distributions that the AG. Moreover a comparison of the utility for using one or another one techniques is make. (Author)

  15. Higher harmonic imaging of tensile plastic deformation in loading and reloading processes by local resonance method

    International Nuclear Information System (INIS)

    Kawashima, Koichiro; Yasui, Hajime

    2015-01-01

    We have imaged plastically deformed region in a 5052 aluminum plate under tensile loading, unloading and reloading processes by using an immersion local resonance method. By transmitting large-amplitude burst wave of which frequency is a through-thickness resonant frequency of the plate, dislocation loops in plastic zone are forced to vibrate. The higher harmonic amplitude excited by the dislocation movement is mapped for the transducer position. The extension of plastic zone under monotonically increased loading, decrease in harmonic amplitude under unloading process and marked extension of plastic zone in reloading up to 0.4% plastic strain are clearly imaged. (author)

  16. Examination of fuel reinsertion strategies for out-of core fuel management

    International Nuclear Information System (INIS)

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

    1986-01-01

    A computer code for determining out-of-core fuel loading strategies in order to minimize levelized fuel cycle cost within constraints has been developed and previously reported by the authors. While past work in this area has dealt with optimizations during equilibrium operating conditions, this work has considered the more realistic conditions of nonequilibrium cycles. The code, called OCEON, seeks to determine a family of economically attractive fuel reload strategies through the optimum selection of feed batch sizes, enrichments, and partially burned fuel reinsertion strategies within operating constraints. This paper presents recent work on expanding the code to allow for different fuel reinsertion options when determining the family of near-optimum fuel reload strategies

  17. Dose distribution, using homogeneous material before the reload of the JS-6500 irradiator

    International Nuclear Information System (INIS)

    Carrasco A, H.

    1991-10-01

    The objective of this report is to determine the dose distribution inside the aluminum containers used for the industrial irradiation, as well as to locate the positions of maximum and minimum doses, before the reloading of the JS-6500 Irradiator. (Author)

  18. Methodology for LOCA analysis and its qualification procedures for PWR reload licensing

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1986-01-01

    The methodology for LOCA analysis developed by FURNAS and its qualification procedure for PWR reload licensing are presented. Digital computer codes developed by NRC and published collectively as the WREM package were modified to get versions that comply to each requirement of Brazilian Licensing Criteria. This metodology is applied to Angra-1 basic case to conclude the qualification process. (Author) [pt

  19. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    Aragones, J.M.

    1977-01-01

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

  20. An application of neural networks and artificial intelligence for in-core fuel management

    International Nuclear Information System (INIS)

    Miller, L.F.; Algutifan, F.; Uhrig, R.E.

    1992-01-01

    This paper reports the feasibility of using expert systems in combination with neural networks and neutronics calculations to improve the efficiency for obtaining optimal candidate reload core designs. The general objectives of this research are as follows: (1) generate a suitable data base and ancillary software for training neural networks that duplicate neutronics calculations. (2) develop a graphical interface with neutronics software and neural networks for manual shuffling of reload cores. (3) construct an expert system for shuffling reload cores with specified rules. (4) develp neural networks that capture the nonlinear behavior of fuel depletion. (5) integrate the neural networks and neutronics software with an expert system to specify reload cores that obtain appropriate figure of merit

  1. New phenomena observed during fuel assemblies testing

    International Nuclear Information System (INIS)

    Tzotcheva, V.

    2001-01-01

    The paper presents a new attempt to explain inexplicable increase of specific activity for some of the fuel assemblies during the fuel tightness testing procedures on Kozloduy NPP. A brief description of established procedure for fuel tightness control is presented in the paper. Special emphasis is given on a hypothesis that explains the fact of existence of deviation in Iodine activity more than usual, which have no reasonable interpretation. The reasons for uniform high Iodine activity for reloaded assemblies, that have kept in the open measuring can for a long time (1-3 hours), is found to be the process of Iodine dissolving in the water and the accelerated process of natural degassing. A proposal to use the 134 Cs and 137 Cs as stand-alone criteria for more precise results is made in respect to increase the reliability of fuel reloading and storage procedures

  2. An optimized BWR fuel lattice for improved fuel utilization

    International Nuclear Information System (INIS)

    Bernander, O.; Helmersson, S.; Schoen, C.G.

    1984-01-01

    Optimization of the BWR fuel lattice has evolved into the water cross concept, termed ''SVEA'', whereby the improved moderation within bundles augments reactivity and thus improves fuel cycle economy. The novel design introduces into the assembly a cruciform and double-walled partition containing nonboiling water, thus forming four subchannels, each of which holds a 4x4 fuel rod bundle. In Scandinavian BWRs - for which commercial SVEA reloads are now scheduled - the reactivity gain is well exploited without adverse impact in other respects. In effect, the water cross design improves both mechanical and thermal-hydraulic performance. Increased average burnup is also promoted through achieving flatter local power distributions. The fuel utilization savings are in the order of 10%, depending on the basis of comparison, e.g. choice of discharge burnup and lattice type. This paper reviews the design considerations and the fuel utilization benefits of the water cross fuel for non-Scandinavian BWRs which have somewhat different core design parameters relative to ASEA-ATOM reactors. For one design proposal, comparisons are made with current standard 8x8 fuel rod bundles as well as with 9x9 type fuel in reactors with symmetric or asymmetric inter-assembly water gaps. The effect on reactivity coefficients and shutdown margin are estimated and an assessment is made of thermal-hydraulic properties. Consideration is also given to a novel and advantageous way of including mixed-oxide fuel in BWR reloads. (author)

  3. Neutronics benchmarks of mixed-oxide fuels using the SCALE/CENTRM sequence

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Fox, P.B.

    2000-01-01

    The purpose of this study is to determine and document the reactor physics parameters (multiplication factors, spatially dependent flux ratios, and spacially dependent reaction rates ) for several distinct sets of problems using two distinct resonance cross-section processing techniques. In SCALE, by default, resonances are processed using NITAWL, which utilizes the Nordheim Integral Treatment. The results produced using this sequence are considered to be the base results. A second set of results are produced by replacing NITAWL with CENTRM/PMC. CENTRM produces point-wise fluxes for a given geometry configuration and set of isotopes. Using these fluxes, PMC produces problem-dependent self-shielding cross sections. Both sequences use ENDF/B-V cross-section data

  4. Fuel elements for LWR power plants

    International Nuclear Information System (INIS)

    Roepenack, H.

    1977-01-01

    About five times more expensive than the fabrication of a fuel element is the enriched uranium contained therein; soon the monthly interest charges for the uranium value of a fuel element reload will account for five percent of the fabrication costs, and much more expensive than all this together can it be if reactor operation has to be interrupted because of damaged elements. Thus, quality assurance comes first. (orig.) [de

  5. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel; Diseno de un nucleo de equilibrio de un reactor tipo BWR basado en un combustible de Torio-Uranio

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Nunez C, A. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

    2003-07-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  6. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  7. Overview of neutronic fuel assembly design and in-core fuel management

    International Nuclear Information System (INIS)

    Porsch, D.; Charlier, A.; Meier, G.; Mougniot, J.C.; Tsuda, K.

    2000-01-01

    The civil and military utilization of nuclear power results in stockpiles of spent fuel and separated plutonium. Recycling of the recovered plutonium in Light Water Reactors (LWR) is currently practiced in Belgium, France, Germany, and Switzerland, in Japan it is in preparation. Modern MOX fuel, with its optimized irradiation and reprocessing behavior, was introduced in 1981. Since then, about 1700 MOX fuel assemblies of different mechanical and neutronic design were irradiated in commercial LWRs and reached fuel assembly averaged exposures of up to 51.000 MWd/t HM. MOX fuel assemblies reloaded in PWR have an average fissile plutonium content of up to 4.8 w/o. For BWR, the average fissile plutonium content in actual reloads is 3.0 w/o. Targets for the MOX fuel assembly design are the compatibility to uranium fuel assemblies with respect to their mechanical fuel rod and fuel assembly design, they should have no impact on the flexibility of the reactor operation, and its reload should be economically feasible. In either cycle independent safety analyses or individually for each designed core it has to be demonstrated that recycling cores meet the same safety criteria as uranium cores. The safety criteria are determined for normal operation and for operational as well as design basis transients. Experience with realized MOX core loadings confirms the reliability of the applied modern design codes. Studies for reloads of advanced MOX assemblies in LWRs demonstrate the feasibility of a future development of the thermal plutonium recycling. New concepts for the utilization of plutonium are under consideration and reveal an attractive potential for further developments on the plutonium exploitation sector. (author)

  8. Insulin-like growth factor-1 receptor in mature osteoblasts is required for periosteal bone formation induced by reloading

    Science.gov (United States)

    Kubota, Takuo; Elalieh, Hashem Z.; Saless, Neema; Fong, Chak; Wang, Yongmei; Babey, Muriel; Cheng, Zhiqiang; Bikle, Daniel D.

    2013-11-01

    Skeletal loading and unloading has a pronounced impact on bone remodeling, a process also regulated by insulin-like growth factor-1 (IGF-1) signaling. Skeletal unloading leads to resistance to the anabolic effect of IGF-1, while reloading after unloading restores responsiveness to IGF-1. However, a direct study of the importance of IGF-1 signaling in the skeletal response to mechanical loading remains to be tested. In this study, we assessed the skeletal response of osteoblast-specific Igf-1 receptor deficient (Igf-1r-/-) mice to unloading and reloading. The mice were hindlimb unloaded for 14 days and then reloaded for 16 days. Igf-1r-/- mice displayed smaller cortical bone and diminished periosteal and endosteal bone formation at baseline. Periosteal and endosteal bone formation decreased with unloading in Igf-1r+/+ mice. However, the recovery of periosteal bone formation with reloading was completely inhibited in Igf-1r-/- mice, although reloading-induced endosteal bone formation was not hampered. These changes in bone formation resulted in the abolishment of the expected increase in total cross-sectional area with reloading in Igf-1r-/- mice compared to the control mice. These results suggest that the Igf-1r in mature osteoblasts has a critical role in periosteal bone formation in the skeletal response to mechanical loading.

  9. Source rack reload of the Tunisian gamma irradiation facility using Monte Carlo method

    International Nuclear Information System (INIS)

    Gharbi, Foued; Kadri, Omrane

    2005-01-01

    This work presents a Monte Carlo study of the cylindrical source rack geometry of the tunisian gamma irradiation facility, using Ge ant code of CERN. The study investigates the question of the reload of the source rack. studied configurations consist on housing four new pencils, two in the upper and two in the lower cylinder of the source rack. global dose rate uniformity inside a ''dummy's' product for the case of routine and non routine irradiation and as function of the product bulk density was calculated for eight hypothetical configurations. the same calculation was also performed for both of the original and the ideal, but not practical configuration. It was shown that hypothetical cases produced dose uniformity variations, according to product density, that were statistically no different than the original and the ideal configurations and that reload procedure can not improve the irradiation quality inside the tunisian facility

  10. Qinshan NPP in-core fuel management improvement

    International Nuclear Information System (INIS)

    Kong Deping; Liao Zejun; Wu Xifeng; Wei Wenbin; Wang Yongming; Li Hua

    2006-01-01

    In the 10-year operation of Qinshan Nuclear Power Plant, the initial designed reloading strategy has been improved step by step based on the operation experiences and the advanced domestic and international fuel management methods. Higher burnup has been achieved and more economic operation gained through the loading pattern improvement and the fuel enrichment increased. The article introduces the in-core fuel management strategy improvement of Qinshan Nuclear Power Plant in its 10-year operation. (authors)

  11. Modular approach to LWR in-core fuel management

    International Nuclear Information System (INIS)

    Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

    1980-01-01

    The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

  12. 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

  13. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

  14. Transition cycle fuel management problems of NPP Krsko

    International Nuclear Information System (INIS)

    Petrovic, B.; Pevec, D.; Smuc, T.; Urli, N.

    1989-01-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  15. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

    Vesely, P.; Borovicka, M.

    2001-01-01

    The paper discusses some aspects of high burnup program implementation at Dukovany NPP and its influence on the fuel cycle costs. Dukovany internal fuel cycle is originally designed as a three years cycle of the Out-In-In fuel reloading patterns. These reloads are not only uneconomical but they additionally increased the radiation load of the reactor pressure vessel due to high neutron leakage typical for Out-In-In loading pattern. To avoid the high neutron leakage from the core a transition to 4-year fuel cycle is started in 1987. The neutron leakage from the core is sequentially decreased by insertion of older fuel assemblies at the core periphery. Other developments in fuel cycle are: 1) increasing of enrichment in control assemblies (3.6% of U-235); 2) improvement in fuel assembly design (reduce the assembly shroud thickness from 2.1 to 1.6 mm); 3) introduction of Zr spacer grid instead of stainless steel; 4) introduction of new type of assembly with profiled enrichment with average value of 3.82%. Due to increased reactivity of the new assemblies the transition to the partial 5-year fuel cycle is required. Typical fuel loading pattern for 3, 3.5, 4 and 5-year cycles are shown in the presented paper. An evaluation of fuel cost is also discussed by using comparative analysis of different fuel cycle options. The analysis shows that introduction of the high burnup program has decrease relative fuel cycle costs

  16. Fuel radial design using Path Relinking

    International Nuclear Information System (INIS)

    Campos S, Y.

    2007-01-01

    The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

  17. A non-algorithmic approach to the In-core-fuel management problem of a PWR core

    International Nuclear Information System (INIS)

    Kimhy, Y.

    1992-03-01

    The primary objective of a commercial nuclear power plant operation is to produce electricity a low cost while satisfying safety constraints imposed on the operating conditions. Design of a fuel reload cycle for the current generation nuclear power plant represents a multistage process with a series of design decisions taken at various time points. Of these stages, reload core design is an important stage, due to its impact on safety and economic plant performance parameters. Overall. performance of the plant during the power production cycle depends on chosen fresh fuel parameters, as well as specific fuel configuration of the reactor core. The motivation to computerize generation and optimization of fuel reload configurations follows from some reasons: first, reload is performed periodically and requires manipulation of a large amount of data. second, in recent years, more complicated fuel loading patterns were developed and implemented following changes in fuel design and/or operational requirements, such as, longer cycles, advanced burnable poison designs, low leakage loading patterns and reduction of irradiation-induced damage of the pressure vessel. An algorithmic approach to the problem was generally adopted. The nature of the reload design process is a 'heuristic' search performed manually by a fuel manager. The knowledge used by the fuel manager is mostly accumulated experience in reactor physics and core calculations. These features of the problem and the inherent disadvantage of the algorithmic method are the main reasons to explore a non-algorithmic approach for solving the reload configuration problem. Several features of the 'solutions space' ( a collection of acceptable final configurations ) are emphasized in this work: 1) the space contain numerous number of entities (> 25) that are distributed un homogeneously, 2) the lack of a monotonic objective function decrease the probability to find an isolated optimum configuration by depth first search or

  18. Experience of safety and performance improvement for fuel handling equipment

    International Nuclear Information System (INIS)

    Gyoon Chang, Sang; Hee Lee, Dae

    2014-01-01

    The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

  19. Licensing and advanced fuel designs

    International Nuclear Information System (INIS)

    Davidson, S.L.; Novendstern, E.H.

    1991-01-01

    For the past 15 years, Westinghouse has been actively involved in the development and licensing of fuel designs that contain major advanced features. These designs include the optimized fuel assembly, The VANTAGE 5 fuel assembly, the VANTAGE 5H, and most recently the VANTAGE+ fuel assembly. Each of these designs was supported by extensive experimental data, safety evaluations, and design efforts and required intensive interaction with the US Nuclear Regulatory Commission (NRC) during the review and approval process. This paper presents a description of the licensing approach and how it was utilized by the utilities to facilitate the licensing applications of the advanced fuel designs for their plants. The licensing approach described in this paper has been successfully applied to four major advanced fuel design changes ∼40 plant-specific applications, and >350 cycle-specific reloads in the past 15 years

  20. In core fuel management optimization by varying the equilibrium cycle average flux shape for batch refuelled reactors

    International Nuclear Information System (INIS)

    Jong, A.J. de.

    1992-12-01

    We suggest a method to overcome this problem of optimization by varying reloading patterns by characterizing each particular reloading pattern by a set of intermediate parameters that are numbers. Plots of the objective function versus the intermediate parameters can be made. When the intermediate parameters represent the reloading patterns in a unique way, the optimum of the objective function can be found by interpolation within such plots and we can find the optimal reloading pattern in terms of intermediate parameters. These have to be transformed backwards to find an optimal reloading pattern. The intermediate parameters are closely related to the time averaged neutron flux shape in the core during an equilibrium cycle. This flux shape is characterized by a set of ratios of the space averaged fluxes in the fuel zones and the space averaged flux in the zone with the fresh fuel elements. An advantage of this choice of intermediate parameters is that it permits analytical calculation of equilibrium cycle fuel densities in the fuel zones for any applied reloading patten characterized by a set of equilibrium cycle average flux ratios and thus, provides analytical calculations of fuel management objective functions. The method is checked for the burnup of one fissile nuclide in a reactor core with the geometry of the PWR at Borssele. For simplicity, neither the conversion of fuel, nor the buildup of fission products were taken into account in this study. Since these phenomena can also be described by the equilibrium cycle average flux ratios, it is likely that this method can be extended to a more realistic method for global in core fuel management optimization. (orig./GL)

  1. 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

  2. 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

  3. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  4. A new coupled system for BWR nuclear fuel management

    International Nuclear Information System (INIS)

    Castillo, A.; Ortiz-Servin, J.J.; Montes-Tadeo, J.L.; Perusquia, R.; Rizos, R.L.M.

    2015-01-01

    In this work, a system to solve four stages of the fuel management problem is showed.The system uses different heuristic techniques to solve each stage of that area, and this problem is solved in a coupled way. Considered problems correspond to the following designs: fuel lattice, fuel assembly, fuel reload and control rod patterns. Even though, each stage of the problem can have its own objective function, the complete problem was solved using a multi-objective function. The solution strategy is to solve each stage of design in an iterative process, taking into account previous results for the next stage, until to achieve a complete solution. The solution strategy to solve the coupled problem is the following: the first solved stage is the fuel lattice design, the second one is fuel assembly design, finally an internal loop between both fuel reload design and control rod pattern design is carried out.For this internal loop, a seed reload using Haling principle is generated. The obtained results showed the advantage to solve the whole problem in a coupled way. (author)

  5. 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)

  6. Multiobjective pressurized water reactor reload core design by nondominated genetic algorithm search

    International Nuclear Information System (INIS)

    Parks, G.T.

    1996-01-01

    The design of pressurized water reactor reload cores is not only a formidable optimization problem but also, in many instances, a multiobjective problem. A genetic algorithm (GA) designed to perform true multiobjective optimization on such problems is described. Genetic algorithms simulate natural evolution. They differ from most optimization techniques by searching from one group of solutions to another, rather than from one solution to another. New solutions are generated by breeding from existing solutions. By selecting better (in a multiobjective sense) solutions as parents more often, the population can be evolved to reveal the trade-off surface between the competing objectives. An example illustrating the effectiveness of this novel method is presented and analyzed. It is found that in solving a reload design problem the algorithm evaluates a similar number of loading patterns to other state-of-the-art methods, but in the process reveals much more information about the nature of the problem being solved. The actual computational cost incurred depends on the core simulator used; the GA itself is code independent

  7. Storage of spent nuclear fuel: The problem of spent nuclear fuel in Bulgaria

    Energy Technology Data Exchange (ETDEWEB)

    Boyadjiev, Z [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Vapirev, E I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1994-12-31

    The practice of spent nuclear fuel (SNF) management in Bulgaria is briefly described and the problems facing the Kozloduy NPP managing staff in finding safe and economically reasonable way for SNF storage are outlined. Taking into account the current situation in the country, the authors recommend a very careful analysis to be performed for the various options before the `deferred decision` to be taken because it concerns approximately 12000 fuel assemblies for a term of 40-50 years. Some recommendations about assessment of different technologies are given. The following requirements in addition to nuclear safety are proposed to be considered: (1) compatibility of possible technologies for transport to reprocessing plants or final disposal preconditioning facilities; (2) minimization of the operations for reloading, especially for reloading under water after intermediate dry storage; (3) participation of Bulgarian companies in the project. 1 tab., 14 refs.

  8. Copenhagen Reloaded

    Energy Technology Data Exchange (ETDEWEB)

    Fink, Helmut [Institut fuer Theoretische Physik, Universitaet Erlangen-Nuernberg (Germany)

    2009-07-01

    The spirit of Copenhagen seems to have lost its convincing power over the past decades, mainly due to its dubious concepts of ''measurement'' and ''observer''. Quantum universalists, on the other side, claim to describe the transition from quantum to classical behaviour in purely quantum-theoretical terms. The quantum logic approach is often considered as corroborating this view, because classical structures can be embedded into quantum structures. Nevertheless, the universalistic turn is misguided: There is a semantic discontinuity between quantum states and classical facts. Decoherence does not solve the measurement problem. Unsharp properties don't either. The concept of classical limit is of course useful, but not for this purpose. Our conclusion is inconvenient: There is a clash of ontologies. The best way out is offered by accepting the legitimacy of both quantum and classical ontology. Classical ontology sets the stage for the quantum play. With this idea in mind, we promote a fresh look at the quantum interpretation debate, from a Neo-Copenhagen perspective, and without ''observers''.

  9. Microcosm reloaded!

    CERN Multimedia

    Stefania Pandolfi

    2015-01-01

    On 20 July, Microcosm reopened its doors to the public. CERN’s guides are already leading many enthusiastic visitors through the new attractions of the popular exhibition. More areas will reopen progressively over the coming weeks before the official inauguration this autumn.   A group of visitors is guided through the new Microcosm exhibit. After six months of extensive refurbishment work, Microcosm is back and open for public visits. As some installations are not yet up and running, CERN’s guides are there to provide visitors with additional explanations. “The two groups I guided were amazed to see real detector components before their eyes, laid out in inviting and informative displays,” says Achintya Rao, member of the CMS collaboration and a CERN guide. “The life-sized CMS mock-up and the CCC panorama are excellent locations to have in-depth conversations with enthusiastic visitors and give them a glimpse into everyday life at CERN.&...

  10. Iliad Reloaded

    DEFF Research Database (Denmark)

    Kesting, Peter; Thiel, Stephanie

    2014-01-01

    based on the historical epos of Homer. However, the negotiation setting is fictitious. In addition, the role of the Hittites slightly twists history. In fact, research assumes that Troy was already under the Hittites’ rule at the time of the Trojan War. However, there is still some uncertainty about......After Paris, Prince of Troy, captured Helena, Menelaus (Helena’s husband, King of Sparta) was thirsting for revenge. He and his brother, Agamemnon (King of Mycenae), forged the Alliance of the Achaeans and jointly went to war against Troy. However, after ten years of siege, the Achaeans realized...... that they could not win the war (aside from Odysseus, who had an idea involving a wooden horse, but people did not think his idea would work). The war is ruining the economy of both parties; further, both are threatened by the Hittites, a hostile kingdom in the East. It is time to negotiate! However, do...

  11. Exopolis reloaded

    NARCIS (Netherlands)

    Altin, Roberta; Minca, Claudio

    2017-01-01

    This paper examines a ‘landscape of power’ in the marginal northeastern corner of Italy, near the Italian-Slovenian border. The landscape is centred around the small town of Gradisca and its highly contested centres for asylum seekers and marked by the concomitant presence of a giant shopping

  12. Reload of cobalt 60 for the J S-6500 irradiator; Recarga de cobalto 60 para el irradiador JS-6500

    Energy Technology Data Exchange (ETDEWEB)

    Torres C, G; Mayoral G, V M

    1991-01-15

    The present work has the purpose to describe the activities of the reloads program of the industrial irradiator J S-6500, elaborated for satisfying part of the demand of services and as a first step, to guide the decision making by the part of the ININ authorities in front of a wide market of this service. (Author)

  13. Transition cycle fuel management problems of NPP Krsko; Problemi gospodarenje gorivom u prijelaznim ciklusima NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B [Institut ' Rudjer Boskovic' , Zagreb (Yugoslavia); Pevec, D [Elektrotehnicki fakultet, Zagreb (Yugoslavia); Smuc, T; Urli, N [Institut ' Rudjer Boskovic' , Zagreb (Yugoslavia)

    1989-07-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  14. Solar cell reloaded; Solarzelle reloaded

    Energy Technology Data Exchange (ETDEWEB)

    Iken, Joern

    2013-06-06

    Who comes up with something special, he may also compete with Chinese. The German-Scandinavian company Innotech Solar extends its solar module production capacity even in the midst of the crisis. Innotech Solar restores damaged cells. For this, the damaged areas are isolated and inactivated. [German] Wer sich etwas Besonderes einfallen laesst, kann auch mit chinesischer Konkurrenz bestehen. Das deutsch-skandinavische Unternehmen Innotech Solar erweitert seine Kapazitaet zur Modulherstellung sogar mitten in der Krise. Das Geschaeftsmodell der Innotech Solar sieht vor, vorgeschaedigte Solarzellen wiederherzustellen. Dafuer werden die schadhaften Stellen isoliert und stillgelegt.

  15. Novel Diagonal Reloading Based Direction of Arrival Estimation in Unknown Non-Uniform Noise

    Directory of Open Access Journals (Sweden)

    Hao Zhou

    2018-01-01

    Full Text Available Nested array can expand the degrees of freedom (DOF from difference coarray perspective, but suffering from the performance degradation of direction of arrival (DOA estimation in unknown non-uniform noise. In this paper, a novel diagonal reloading (DR based DOA estimation algorithm is proposed using a recently developed nested MIMO array. The elements in the main diagonal of the sample covariance matrix are eliminated; next the smallest MN-K eigenvalues of the revised matrix are obtained and averaged to estimate the sum value of the signal power. Further the estimated sum value is filled into the main diagonal of the revised matrix for estimating the signal covariance matrix. In this case, the negative effect of noise is eliminated without losing the useful information of the signal matrix. Besides, the degrees of freedom are expanded obviously, resulting in the performance improvement. Several simulations are conducted to demonstrate the effectiveness of the proposed algorithm.

  16. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This is the `94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author).

  17. Development of the advanced PHWR technology -Verification tests for CANDU advanced fuel-

    International Nuclear Information System (INIS)

    Jung, Jang Hwan; Suk, Hoh Chun; Jung, Moon Kee; Oh, Duk Joo; Park, Joo Hwan; Shim, Kee Sub; Jang, Suk Kyoo; Jung, Heung Joon; Park, Jin Suk; Jung, Seung Hoh; Jun, Ji Soo; Lee, Yung Wook; Jung, Chang Joon; Byun, Taek Sang; Park, Kwang Suk; Kim, Bok Deuk; Min, Kyung Hoh

    1995-07-01

    This is the '94 annual report of the CANDU advanced fuel verification test project. This report describes the out-of pile hydraulic tests at CANDU-hot test loop for verification of CANFLEX fuel bundle. It is also describes the reactor thermal-hydraulic analysis for thermal margin and flow stability. The contents in this report are as follows; (1) Out-of pile hydraulic tests for verification of CANFLEX fuel bundle. (a) Pressure drop tests at reactor operation condition (b) Strength test during reload at static condition (c) Impact test during reload at impact load condition (d) Endurance test for verification of fuel integrity during life time (2) Reactor thermal-hydraulic analysis with CANFLEX fuel bundle. (a) Critical channel power sensitivity analysis (b) CANDU-6 channel flow analysis (c) Flow instability analysis. 61 figs, 29 tabs, 21 refs. (Author)

  18. Fuel Management at the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

    2011-07-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

  19. Influence of short-term unweighing and reloading on running kinetics and muscle activity.

    Science.gov (United States)

    Sainton, Patrick; Nicol, Caroline; Cabri, Jan; Barthelemy-Montfort, Joëlle; Berton, Eric; Chavet, Pascale

    2015-05-01

    In running, body weight reduction is reported to result in decreased lower limb muscle activity with no change in the global activation pattern (Liebenberg et al. in J Sports Sci 29:207-214). Our study examined the acute effects on running mechanics and lower limb muscle activity of short-term unweighing and reloading conditions while running on a treadmill with a lower body positive pressure (LBPP) device. Eleven healthy males performed two randomized running series of 9 min at preferred speed. Each series included three successive running conditions of 3 min [at 100 % body weight (BW), 60 or 80 % BW, and 100 % BW]. Vertical ground reaction force and center of mass accelerations were analyzed together with surface EMG activity recorded from six major muscles of the left lower limb for the first and last 30 s of each running condition. Effort sensation and mean heart rate were also recorded. In both running series, the unloaded running pattern was characterized by a lower step frequency (due to increased flight time with no change in contact time), lower impact and active force peaks, and also by reduced loading rate and push-off impulse. Amplitude of muscle activity overall decreased, but pre-contact and braking phase extensor muscle activity did not change, whereas it was reduced during the subsequent push-off phase. The combined neuro-mechanical changes suggest that LBPP technology provides runners with an efficient support during the stride. The after-effects recorded after reloading highlight the fact that 3 min of unweighing may be sufficient for updating the running pattern.

  20. Taiwan Power Company's power distribution analysis and fuel thermal margin verification methods for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, P.H.

    1995-01-01

    Taiwan Power Company's (TPC's) power distribution analysis and fuel thermal margin verification methods for pressurized water reactors (PWRs) are examined. The TPC and the Institute of Nuclear Energy Research started a joint 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, these methods were developed to allow TPC to independently perform verifications of the local power density and departure from nucleate boiling design bases, which are required by the reload safety evaluation for the Maanshan PWR plant. The computer codes utilized were extensively validated for the intended applications. Sample calculations were performed for up to six reload cycles of the Maanshan plant, and the results were found to be quite consistent with the vendor's calculational results

  1. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  2. The continual fuel management modification in Qinshan project II

    International Nuclear Information System (INIS)

    Ye Guodong; Pan Zefei; Zhang Xingtian

    2010-01-01

    The fuel management strategy is the basis of the nuclear power plants. The performance of the fuel management strategy affects the plants' safety and economy indicators directly. The paper summarizes all the modifications on the fuel management work in Qinshan Project II since the plant was established. It includes the surveillance system of physics tests, fetching in high performance fuel assemblies, reloading pattern optimization, and the modifications of the final safety analysis report. At the same time, it evaluates the benefit of the modifications in the few years. The experience in this paper is much helpful and could be implemented on the same type plants. (authors)

  3. The development of fuel elements for boiling water reactors

    International Nuclear Information System (INIS)

    Holzer, R.; Kilian, P.

    1984-01-01

    The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

  4. Method to generate the first design of the reload pattern to be used with the Presto-B code in the simulation of the CNLV U-1 reactor

    International Nuclear Information System (INIS)

    Montes T, J.L.; Cortes C, C.C.

    1992-08-01

    This guide is applied for the reload pattern's formation for mirror symmetry of a core room and in accordance with the Control Cell core technique (of the english Control Cell Core - CCC) for the PRESTO-B code. (Author)

  5. System to solve three designs of the fuel management

    International Nuclear Information System (INIS)

    Castillo M, J. A.; Ortiz S, J. J.; Montes T, J. L.; Perusquia del C, R.; Marinez R, R.

    2015-09-01

    In this paper preliminary results are presented, obtained with the development of a computer system that resolves three stages of the nuclear fuel management, which are: the axial and radial designs of fuel, as well as the design of nuclear fuel reloads. The novelty of the system is that the solution is obtained solving the 3 mentioned stages, in coupled form. For this, heuristic techniques are used for each stage, in each one of these has a function objective that is applied to particular problems, but in all cases the obtained partial results are used as input data for the next stage. The heuristic techniques that were used to solve the coupled problem are: tabu search, neural networks and a hybrid between the scatter search and path re linking. The system applies an iterative process from the design of a fuel cell to the reload design, since are preliminary results the reload is designed using the operation strategy Haling type. In each one of the stages nuclear parameters inherent to the design are monitored. The results so far show the advantage of solving the problem in a coupled manner, even when a large amount of computer resources is used. (Author)

  6. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  7. Fuel management approach in IRIS Reactor

    International Nuclear Information System (INIS)

    Petrovic, B.; Franceschini, F.

    2004-01-01

    This paper provides an overview of fuel management approach employed in IRIS (International Reactor Innovative and Secure). It introduces the initial, rather ambitious, fuel management goals and discusses their evolution that reflected the fast pace of progress of the overall project. The updated objectives rely on using currently licensed fuel technology, thus enabling near-term deployment of IRIS, while still providing improved fuel utilization. The paper focuses on the reference core design and fuel management strategy that is considered in pre-application licensing, which enables extended cycle of three to four years. The extended cycle reduces maintenance outage time and increases capacity factor, thus reducing the cost of electricity. Approaches to achieving this goal are discussed, including use of different reloading strategies. Additional fuel management options, which are not part of the licensing process, but are pursued as long-term research for possible future implementation, are presented as well. (Author)

  8. Experience of TVSA fuel implementation at Kozloduy NPP

    International Nuclear Information System (INIS)

    Kamenov, K.; Kamenov, AI.; Hristov, D.

    2011-01-01

    The base design of the Russian fuel assemblies TVSA have been under operation at Kozloduy NPP WWER-1000 reactors since 2004. The old type fuel assemblies TVS-M were gradually substituted till 2008. The TVSA assembly distinguishes itself with much stronger construction. As a burnable absorber it has a mixture of uranium and uniformly distributed Gd in 6 or more fuel rods. This enables to increase the safety and effectiveness of fuel cycles. The experience gained during TVSA fuel implementation on units 5 and 6 and KASKAD code package validation was presented at the eightieth International conference on WWER 'Fuel performance, modelling and experimental support in 2009'. Additional information about TVSA fuel implementation at Kozloduy NPP WWER-1000 units in a 4-year fuel cycle with 42 and 48 fresh fuel assemblies reloading scheme is presented in the paper. (Authors)

  9. Review of primary spaceflight-induced and secondary reloading-induced changes in slow antigravity muscles of rats

    Science.gov (United States)

    Riley, D. A.

    We have examined the light and electron microscopic properties of hindlimb muscles of rats flown in space for 1-2 weeks on Cosmos biosatellite flights 1887 and 2044 and Space Shuttle missions Spacelab-3, Spacelab Life Sciences-1 and Spacelab Life Sciences-2. Tissues were obtained both inflight and postflight permitting definition of primary microgravity-induced changes and secondary reentry and gravity reloading-induced alterations. Spaceflight causes atrophy and expression of fast fiber characteristics in slow antigravity muscles. The stresses of reentry and reloading reveal that atrophic muscles show increased susceptibility to interstitial edema and ischemic-anoxic necrosis as well as muscle fiber tearing with disruption of contractile proteins. These results demonstrate that the effects of spaceflight on skeletal muscle are multifaceted, and major changes occur both inflight and following return to Earth's gravity.

  10. Stakes and Solutions for current and up-coming Licensing Challenges in PWR and BWR Reload and Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tiving, F.; Opel, S.

    2014-07-01

    Regulatory requirements for reloads and safety analyses are evolving: New safety criteria, requests for enlarged qualification databases, statistical applications, uncertainty propagation... In order to address these challenges and access more predictable licensing processes, AREVA implements a consistent code and methodology suite for PWR and BWR core design and safety analysis, based on a first principles modeling with an extremely broad international verification and validation data base. (Author)

  11. The Reactor Analysis Support Package (RASP): Volume 10, Guidelines for developing a reload licensing capability:Final report

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1988-08-01

    The EPRI Reactor Analysis Support Package (RASP) consists of computer codes and documentation for calculating core performance and plant transients. This report was written to help a utility to use these tools properly for reload design and safety evaluations. The emphasis is on the steps that the utility can take to develop a methodology that is approved by the Nuclear Regulatory Commission for reload licensing submittals. The report treats both the planning and the implementation of this type of calculational capability. With regard to planning, there is discussion defining objectives, resource requirements, organization and scheduling. In order to help the engineering staff implement the plan there is discussion of the development of a methodology for event analysis, qualification of the methods, and the writing of design control procedures and topical reports. The experience of utilities, and especially of GPU Nuclear Corporation (GPUN), in developing a reload licensing capability is cited throughout the report and extracts from GPUN design control procedures are included in the appendices. 16 refs., 23 figs., 9 tabs

  12. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

  13. Development and preliminary analyses of material balance evaluation model in nuclear fuel cycle

    International Nuclear Information System (INIS)

    Matsumura, Tetsuo

    1994-01-01

    Material balance evaluation model in nuclear fuel cycle has been developed using ORIGEN-2 code as basic engine. This model has feature of: It can treat more than 1000 nuclides including minor actinides and fission products. It has flexibility of modeling and graph output using a engineering work station. I made preliminary calculation of LWR fuel high burnup effect (reloading fuel average burnup of 60 GWd/t) on nuclear fuel cycle. The preliminary calculation shows LWR fuel high burnup has much effect on Japanese Pu balance problem. (author)

  14. A neutronics study of LEU fuel options for the HFR-Petten

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.

    1985-01-01

    The standard HEU fuel cycle characteristics are compared with those of several different LEU fuel cycles in the new vessel configuration. The primary design goals were to provide similar reactivity performance and neutron flux profiles with a minimal increase in 235 U loading. The fuel cycle advantages of Cd burnable absorbers over 10 B are presented. The LEU fuel cycle requirements were calculated also for an extended 32-day cycle and for a reload batch size reduction from six to five standard elements for the standard 26-day cycle. The effects of typical in-core experiments upon neutron flux profiles and fuel loading requirements are also presented. (author)

  15. New approaches of the potential field for QPSO algorithm applied to nuclear reactor reload problem

    International Nuclear Information System (INIS)

    Nicolau, Andressa dos Santos; Schirru, Roberto

    2015-01-01

    Recently quantum-inspired version of the Particle Swarm Optimization (PSO) algorithm, Quantum Particle Swarm Optimization (QPSO) was proposed. The QPSO algorithm permits all particles to have a quantum behavior, where some sort of 'quantum motion' is imposed in the search process. When the QPSO is tested against a set of benchmarking functions, it showed superior performances as compared to classical PSO. The QPSO outperforms the classical one most of the time in convergence speed and achieves better levels for the fitness functions. The great advantage of QPSO algorithm is that it uses only one parameter control. The critical step or QPSO algorithm is the choice of suitable attractive potential field that can guarantee bound states for the particles moving in the quantum environment. In this article, one version of QPSO algorithm was tested with two types of potential well: delta-potential well harmonic oscillator. The main goal of this study is to show with of the potential field is the most suitable for use in QPSO in a solution of the Nuclear Reactor Reload Optimization Problem, especially in the cycle 7 of a Brazilian Nuclear Power Plant. All result were compared with the performance of its classical counterpart of the literature and shows that QPSO algorithm are well situated among the best alternatives for dealing with hard optimization problems, such as NRROP. (author)

  16. Conceptual development of a complete LWR reload design methodology based on generalized perturbation theory

    International Nuclear Information System (INIS)

    White, J.R.

    1986-01-01

    A new approach for the physics design and analysis of LWR reload cores is developed and demonstrated through several practical applications. The new design philosophy uses first- and second-order response derivatives to predict the important reactor performance characteristics (power peaking, reactivity coefficients, etc.) for any number of possible material configurations (assembly shuffling and burnable poison loadings). The response derivatives are computed using generalized perturbation theory (GPT) techniques. This report describes in detail an idealized GPT-based design system. The idealized system would contain individual modules to generate the required first-order and higher-order sensitivity data. It would also contain at least two major application codes; one for core design optimization and the other for evaluation of several safety parameters of interest in off-normal situations. This ideal system would be fully automated, user-friendly, and quite flexible in its ability to provide a variety of design and analysis capabilities. Information gained form these three studies gives a good foundation for the development of a complete integrated design package

  17. New approaches of the potential field for QPSO algorithm applied to nuclear reactor reload problem

    Energy Technology Data Exchange (ETDEWEB)

    Nicolau, Andressa dos Santos; Schirru, Roberto, E-mail: andressa@lmp.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Recently quantum-inspired version of the Particle Swarm Optimization (PSO) algorithm, Quantum Particle Swarm Optimization (QPSO) was proposed. The QPSO algorithm permits all particles to have a quantum behavior, where some sort of 'quantum motion' is imposed in the search process. When the QPSO is tested against a set of benchmarking functions, it showed superior performances as compared to classical PSO. The QPSO outperforms the classical one most of the time in convergence speed and achieves better levels for the fitness functions. The great advantage of QPSO algorithm is that it uses only one parameter control. The critical step or QPSO algorithm is the choice of suitable attractive potential field that can guarantee bound states for the particles moving in the quantum environment. In this article, one version of QPSO algorithm was tested with two types of potential well: delta-potential well harmonic oscillator. The main goal of this study is to show with of the potential field is the most suitable for use in QPSO in a solution of the Nuclear Reactor Reload Optimization Problem, especially in the cycle 7 of a Brazilian Nuclear Power Plant. All result were compared with the performance of its classical counterpart of the literature and shows that QPSO algorithm are well situated among the best alternatives for dealing with hard optimization problems, such as NRROP. (author)

  18. Possible involvement of 12-lipoxygenase activation in glucose-deprivation/reload-treated neurons.

    Science.gov (United States)

    Nagasawa, Kazuki; Kakuda, Taichi; Higashi, Youichirou; Fujimoto, Sadaki

    2007-12-18

    The aim of this study was to clarify whether 12-lipoxygenase (12-LOX) activation was involved in reactive oxygen species (ROS) generation, extensive poly(ADP-ribose) polymerase (PARP) activation and neuronal death induced by glucose-deprivation, followed by glucose-reload (GD/R). The decrease of neuronal viability and accumulation of poly(ADP-ribose) induced by GD/R were prevented 3-aminobenzamide, a representative PARP inhibitor, demonstrating this treatment protocol caused the same oxidative stress with the previously reported one. The PARP activation, ROS generation and decrease of neuron viability induced by GD/R treatment were almost completely abolished by an extracellular zinc chelator, CaEDTA. p47(phox), a cytosolic component of NADPH oxidase was translocated the membrane fraction by GD/R, indicating its activation, but it did not generate detectable ROS. Surprisingly, pharmacological inhibition of NADPH oxidase with apocynin and AEBSF further decreased the decreased neuron viability induced by GD/R. On the other hand, AA861, a 12-LOX inhibitor, prevented ROS generation and decrease of neuron viability caused by GD/R. Interestingly, an antioxidant, N-acetyl-l-cysteine rescued the neurons from GD/R-induced oxidative stress, implying effectiveness of antioxidant administration. These findings suggested that activation of 12-LOX, but not NADPH oxidase, following to zinc release might play an important role in ROS generation and decrease of viability in GD/R-treated neurons.

  19. Integrated planning for a fuel industry with emphasis on minimum size to fabricate own fuel

    International Nuclear Information System (INIS)

    Kondal Rao, N.; Katiyar, H.C.; Rajendran, R.; Sinha, K.K.; Swaminathan, N.; Subramanyam, R.B.; Pande, B.P.; Krishnan, T.S.; Agarwala, G.C.; Chandramouli, V.A.

    1977-01-01

    The Indian nuclear energy programme is based on the utilization of indigenous resources for the economic generation of power, developing its own know-how. In order to gain time, the first nuclear power station at Tarapur is a turn-key job based on enriched uranium fuel. Taking into consideration the established resources of uranium and thorium in the country, a strategy for nuclear power programme has been drawn up. The first phase is based on natural uranium fuel, the second phase on the recycle of plutonium and conversion of thorium and the third phase is the breeder system based on utilization of U 233 and conversion of thorium. This programme is specially significant for India in view of its vast resources of thorium. After the experience and confidence gained with the manufacture of metallic uranium fuel for the research reactors and about 40 tonnes of fuel for the initial loading of the Rajasthan Reactor, the fuel manufacturing programme within the country has been implemented to meet the entire initial and reload fuel requirements. The plant capacities are small compared to similar activities in developed countries. Further, by planning for an integrated fuel and component manufacturing complex, any draw-back in smaller scale of some of the operations is off-set. At the Nuclear Fuel Complex, set up on the above principles, production plants are in operation for the manufacture of reload fuel for the 400 MW Tarapur station, natural uranium oxide fuel, various zircaloy components such as fuel sheaths, pressure tubes, calandria tubes, channels and various other zircaloy components. Provisions have been made to expand the production facilities as the demand for reload fuel grows. With the facilities provided, the production programme can be diversified to take up the production of fast breeder reactor components of stainless steel and also the blanket thorium elements. The unitary control of all aspects of the manufacture and quality control of different types

  20. FAILED FUEL DISPOSITION STUDY

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.

    2004-12-20

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  1. FAILED FUEL DISPOSITION STUDY

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2004-01-01

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  2. Procedure and apparatus for measuring the radial gap between fuel and surrounding cladding in a fuel rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Olshausen, K.D.

    1976-01-01

    A device is described for measuring non-destructively the annular fuel-cladding gap in an irradiated or fresh fuel rod. The principle applied is that a force is applied to an arm which presses the cladding diametrically, thus deforming it until it touches the fuel pellet. By presenting the values of the force applied and the deformation produced on an XY recorder, the width of the gap is obtained. Alternatively the gap width may be obtained digitally. Since the gap is so small that the deformation is within the elastic range, the fuel rod may be reloaded in the reactor for further irradiation. (JIW)

  3. Investigation of the effects of process sequence on the contact resistance characteristics of coated metallic bipolar plates for polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Turan, Cabir; Cora, Ömer Necati; Koç, Muammer

    2013-12-01

    In this study, results of an investigation on the effects of manufacturing and coating process sequence on the contact resistance (ICR) of metallic bipolar plates (BPP) for polymer electrolyte membrane fuel cells (PEMFCs) are presented. Firstly, uncoated stainless steel 316L blanks were formed into BPP through hydroforming and stamping processes. Then, these formed BPP samples were coated with three different PVD coatings (CrN, TiN and ZrN) at three different thicknesses (0.1, 0.5 and 1 μm). Secondly, blanks of the same alloy were coated first with the same coatings, thickness and technique; then, they were formed into BPPs of the same shape and dimensions using the manufacturing methods as in the first group. Finally, these two groups of BPP samples were tested for their ICR to reveal the effect of process sequence. ICR tests were also conducted on the BPP plates both before and after exposure to corrosion to disclose the effect of corrosion on ICR. Coated-then-formed BPP samples exhibited similar or even better ICR performance than formed-then-coated BPP samples. Thus, manufacturing of coated blanks can be concluded to be more favorable and worth further investigation in quest of making cost effective BPPs for mass production of PEMFC.

  4. Fuel radial design using Path Relinking; Diseno radial de combustible usando Path Relinking

    Energy Technology Data Exchange (ETDEWEB)

    Campos S, Y. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2007-07-01

    The present work shows the obtained results when implementing the combinatory optimization technique well-known as Path Re linking (Re-linkage of Trajectories), to the problem of the radial design of nuclear fuel assemblies, for boiling water reactors (BWR Boiling Water Reactor by its initials in English), this type of reactors is those that are used in the Laguna Verde Nucleo electric Central, Veracruz. As in any other electric power generation plant of that make use of some fuel to produce heat and that it needs each certain time (from 12 to 14 months) to make a supply of the same one, because this it wears away or it burns, in the nucleolectric plants to this activity is denominated fuel reload. In this reload different activities intervene, among those which its highlight the radial and axial designs of fuel assemblies, the patterns of control rods and the multi cycles study, each one of these stages with their own complexity. This work was limited to study in independent form the radial design, without considering the other activities. These phases are basic for the fuel reload design and of reactor operation strategies. (Author)

  5. Alterations in the muscle force transfer apparatus in aged rats during unloading and reloading: Impact of microRNA-31.

    Science.gov (United States)

    Hughes, David C; Marcotte, George R; Baehr, Leslie M; West, Daniel W D; Marshall, Andrea G; Ebert, Scott M; Davidyan, Arik; Adams, Christopher M; Bodine, Sue C; Baar, Keith

    2018-05-03

    Force transfer is integral for maintaining skeletal muscle structure and function. One important component is dystrophin. There is limited understanding of how force transfer is impacted by age and loading. Here, we investigate the force transfer apparatus in muscles of adult and old rats exposed to periods of disuse and reloading. Our results demonstrate an increase in dystrophin protein during the reloading phase in the adult TA muscle that is delayed in old. The consequence of this delay is an increased susceptibility towards contraction-induced muscle injury. Central to the lack of dystrophin protein is an increase in miR-31, a microRNA that inhibits dystrophin translation. In vivo electroporation with a miR-31 sponge led to increased dystrophin protein and decreased contraction-induced muscle injury in old skeletal muscle. Overall, our results detail the importance of the force transfer apparatus and provide new mechanisms for contraction-induced injury in aging skeletal muscle. In healthy muscle, the dystrophin-associated glycoprotein (DGC) and integrin/focal adhesion complexes, intermediate filaments, and Z-line proteins transmit force from the contractile proteins to the extracellular matrix. How loading and age affect these proteins is poorly understood. The experiments reported here sought to determine the effect of aging on the force transfer apparatus following muscle unloading and reloading. Adult (9 months) and old (29 months) rats were subjected to 14 days hindlimb unloading (HU) and 1, 3, 7 and 14 days of reloading (REL). The DGC complex, intermediate filament and z-line protein and mRNA levels, as well as dystrophin-targeting miRNAs (miR-31, -146b and -374) were examined in the tibialis anterior (TA) and medial gastrocnemius (MG) muscles at both ages. There was a significant increase in dystrophin protein levels (2.79-fold) upon 3 days of reloading in the adult TA muscle that did not occur in the old rats (p ≤ 0.05), and the rise in

  6. Global Nuclear Fuel launches GNF{sub 3} and NSF: The most reliable BWR fuel just got better

    Energy Technology Data Exchange (ETDEWEB)

    Cantonwine, P.; Schneider, R.; Hunt, B.

    2015-11-01

    Bases on evolutionary design changes and advanced technology developed by Global Nuclear Fuel (GNF), the GNF3 fuel assembly is designed to offer customers with improved fuel economics, increased performance and flexibility in operation while maintaining the superior reliability of GNF2, the most reliable design in GNFs history. In addition to improved fuel utilization and performance, GNF3 is designed and manufactured to be more resistant to debris capture, to eliminate channel control blade interference concerns, and to exhibit to best available corrosion resistance of any boiling water reactor fuel. While delivering fuel cycle savings and reliability benefits with GNF3, GNF maintains a similar licensing and operating basis to GNF2, thereby minimizing fuel transition risks. GNF3 is available in lead use assembly quantities to customers today. Eight GNF3 lead use assemblies are in operation at two utilities in the USA GNF3 is scheduled to be available for full reloads in 2018. (Author)

  7. Advances in ultrasonic fuel cleaning

    International Nuclear Information System (INIS)

    Blok, J.; Frattini, P.; Moser, T.

    2002-01-01

    The economics of electric generation is requiring PWR plant operators to consider higher fuel duty and longer cycles. As a result, sub-cooled nucleate boiling is now an accepted occurrence in the upper spans of aggressively driven PWR cores. Thermodynamic and hydraulic factors determine that the boiling surfaces of the fuel favor deposition of corrosion products. Thus, the deposits on high-duty fuel tend to be axially distributed in an inhomogeneous manner. Axial offset anomaly (AOA) is the result of axially non-homogeneous distribution of boron compounds in these axially variable fuel deposits. Besides their axial asymmetry, fuel deposits in boiling cores tend to be qualitatively different from deposits on non-boiling fuel. Thus, deposits on moderate-duty PWR fuel are generally iron rich, predominating in nickel ferrites. Deposits on cores with high boiling duty, on the other hand, tend to be rich in nickel, with sizeable fractions of NiO or elemental nickel. Other unexpected compounds such as m-ZrO 2 and Ni-Fe oxy-borates have been found in significant quantity in deposits on boiling cores. This paper describes the ultrasonic fuel cleaning technology developed by EPRI. Data will be presented to confirm that the method is effective for removing fuel deposits from both high-duty and normal-duty fuel. The report will describe full-core fuel cleaning using the EPRI technology for Callaway Cycle 12 reload fuel. The favorable impact of fuel cleaning on Cycle 12 AOA performance will also be presented. (authors)

  8. Womanism and After: A Theatrical Justification for African Women’s Radical Response to Subjugation in Reloaded

    Directory of Open Access Journals (Sweden)

    Sola Emmanuel Owonibi

    2016-07-01

    Full Text Available The inculcation and transmission of socio-cultural, ideological and moral expectations of every society are is as much the role of individuals and institutions of that society as the available oral or written records of the society. With the advent of modernity, the mass media have come to play crucial roles in the socialization and conditioning of members of the society to accepted or expected roles and behaviour. The theatre has come to be very relevant in this regard. Diverse thematic preoccupations have actually authenticated the social relevance of theatre and the home video, especially in the Nigerian Nollywood industry. The focus on inter-personal relationship is particularly remarkable. Some Nollywood movies are particularly exemplary in their deconstruction of the man/woman relationship in the African society. This paper studied ‘Reloaded’, a Nigerian Nollywood movie.  The choice of the movie, Reloaded for this paper is informed by it radical departure from the African feminist tradition of womanism which tends to reject a radical response by women to their subjugation, and rather favours a sort of mild – even passive - dialogic synergy with men. This advocacy for complementarity, as we can see in a movie like Reloaded, has not brought the much-desired solution to women subjugation. This revelation is much more in consonance with reality; the reality that response to issues is spontaneous and universally natural to individuals rather than being unifocal. Using the sociological approach and adopting a critical analysis method, this study finds out that reality in the Nigerian society has shown that, in many cases, passivity, docility and persuasion have failed to bring about desired results where corrective retaliation has done the magic. The man/woman relationship is not an exception, as it is revealed in Reloaded.

  9. Performance Evaluation and Suggestion of Upgraded Fuel Handling Equipment for Operating OPR1000

    International Nuclear Information System (INIS)

    Chang, Sang Gyoon; Hwang, Jeung Ki; Choi, Taek Sang; Na, Eun Seok; Lee, Myung Lyul; Baek, Seung Jin; Kim, Man Su; Kunik, Jack

    2011-01-01

    The purpose of this study is to evaluate the performance of upgraded FHE (Fuel Handling Equipment) for operating OPR 1000 (Optimized Power Reactor) by using data measured during the fuel reloading, and make some suggestions on enhancing the performance of the FHE. The fuel handling equipment, which serves critical processes in the refueling outage, has been upgraded to increase and improve the operational availability of the plant. The evaluation and suggestion of this study can be a beneficial tool related to the performance of the fuel handling equipment

  10. Methods for estimating the reliability of the RBMK fuel assemblies and elements

    International Nuclear Information System (INIS)

    Klemin, A.I.; Sitkarev, A.G.

    1985-01-01

    Applied non-parametric methods for calculation of point and interval estimations for the basic nomenclature of reliability factors for the RBMK fuel assemblies and elements are described. As the fuel assembly and element reliability factors, the average lifetime is considered at a preset operating time up to unloading due to fuel burnout as well as the average lifetime at the reactor transient operation and at the steady-state fuel reloading mode of reactor operation. The formulae obtained are included into the special standardized engineering documentation

  11. Prediction of the local power factor in BWR fuel cells by means of a multilayer neural network

    International Nuclear Information System (INIS)

    Montes, J.L.; Ortiz, J.J.; Perusquia C, R.; Francois, J.L.; Martin del Campo M, C.

    2007-01-01

    To the beginning of a new operation cycle in a BWR reactor the reactivity of this it increases by means of the introduction of fresh fuel, the one denominated reload fuel. The problem of the definition of the characteristics of this reload fuel represents a combinatory optimization problem that requires significantly a great quantity of CPU time for their determination. This situation has motivated to study the possibility to substitute the Helios code, the one which is used to generate the new cells of the reload fuel parameters, by an artificial neuronal network, with the purpose of predicting the parameters of the fuel reload cell of a BWR reactor. In this work the results of the one training of a multilayer neuronal net that can predict the local power factor (LPPF) in such fuel cells are presented. The prediction of the LPPF is carried out in those condition of beginning of the life of the cell (0.0 MWD/T, to 40% of holes in the one moderator, temperature of 793 K in the fuel and a moderator temperature of 560 K. The cells considered in the present study consist of an arrangement of 10x10 bars, of those which 92 contains U 235 , some of these bars also contain a concentration of Gd 2 O 3 and 8 of them contain only water. The axial location inside the one assembles of recharge of these cells it is exactly up of the cells that contain natural uranium in the base of the reactor core. The training of the neuronal net is carried out by means of a retro-propagation algorithm that uses a space of training formed starting from previous evaluations of cells by means of the Helios code. They are also presented the results of the application of the neuronal net found for the prediction of the LPPF of some cells used in the real operation of the Unit One of the Laguna Verde Nuclear Power station. (Author)

  12. Out-of-core fuel cycle optimization for nonequilibrium cycles

    International Nuclear Information System (INIS)

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

    1988-01-01

    A methodology has been developed for determining the family of near-optimum fuel management schemes that minimize the levelized fuel cycle costs of a light water reactor over a multicycle planning horizon. Feed batch enrichments and sizes, burned batches to reinsert, and burnable poison loadings are determined for each cycle in the planning horizon. Flexibility in the methodology includes the capability to assess the economic benefits of various partially burned bath reload strategies as well as the effects of using split feed enrichments and enrichment palettes. Constraint limitations are imposed on feed enrichments, discharge burnups, moderator temperature coefficient, and cycle energy requirements

  13. The evolution of the fuel cycle in the Dukovany NPP

    Energy Technology Data Exchange (ETDEWEB)

    Bajgl, J [Jaderna Elektrarna, Dukovany (Czech Republic)

    1994-12-31

    The ten-year operational experience of four WWER-440 units in Dukovany NPP with a total number of 35 cycles is outlined. The strategy of fuel reloading has been changed from out-in schemes to low-leakage patterns. The linear pin power limitation will be introduced. The main physical limits and conditions for the NPP operation are listed. The main goal is to go to a full 4-year fuel cycle in which the burnup will be about 40 Mwd/kg U. 6 tabs.

  14. Mutual influences of reactor operation and fuel cycle management

    International Nuclear Information System (INIS)

    Lewiner, C.; Schaerer, R.

    1989-01-01

    OPEN (Organisation des Producteurs d'Energie Nucleaire) comprises the electricity producers from seven European countries which now operate or intend to operate nuclear power plants. Its activities include the study of technical, economic and legal subjects related to nuclear electricity. A continuous analysis of the fuel cycle market has been pursued within OPEN for almost 15 years. For the past few years, OPEN has also been concerned with the subject of fuel management in the reactors operated by its members. The purpose of this effort was to obtain an overall picture of possible fuel improvements and to evaluate the effects, in particular the economic ones, of diverse fuel reload managements and of reprocessed uranium and plutonium recycling. The conclusions of this study are as follows: Increase in burn-ups produces notable savings in electricity generating costs. It also permits adaptation of fuel loading mode to the desirable irradiation campaign length. This allows for better management of the country's overall means of electricity generation (nuclear, fossil-fuelled or hydro plants), and adjustment to the electrical demand. These new reload schemes have various impacts on natural uranium consumption and enrichment, but, above all, they affect directly all fuel cycle operations linked to the number of assemblies (fabrication, reprocessing, etc.). 6 figs

  15. Improved fuel design economics - a new evaluation tool

    International Nuclear Information System (INIS)

    Aboudara, J.L.

    1987-01-01

    Advanced fuel design technology is now beginning to be implemented with new reload regions for large pressurized water reactors. Until recently there has not been an integrated computer modeling product that would allow easy assessment of the economics of various advanced fuel design alternatives now available to utilities. The Fuel Cycle Scoping System (FCSS) was designed to fill this need. The FCSS is a personnel computer (PC) software package that is used to evaluate alternative strategies for supplying and using nuclear fuel in nuclear power reactors. The FCSS is an extremely flexible package that permits evaluation of in-core and out-of-core fuel management strategy options. For each strategy option, unit and reactor operating assumptions and assumptions for uranium supply, conversion, enrichment, fabrication, and spent fuel disposal can be made

  16. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  17. Quadratic reactivity fuel cycle model

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1985-01-01

    For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau 2 as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau 2 in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper

  18. In-core fuel management: New challenges

    International Nuclear Information System (INIS)

    Kolmayer, A.; Vallee, A.; Mondot, J.

    1992-01-01

    Experience accumulated by pressurized water reactor (PWR) utilities allows them to improve their strategies in the use of eventual margins to core design limits. They are used for nuclear steam supply system (NSSS) power upgrading, to improve operating margins, or to adapt fuel management to specific objectives. As a result, in-core fuel management strategies have become very diverse: UO 2 or mixed-oxide loading, out-in or in-out fuel loading patterns, extended or annual cycle lengths with margins on design limits such as moderator temperature coefficients, boron concentrations, or peaking factors. Perspectives also appear concerning use of existing plutonium stocks or actinide incineration. Burnable poisons are most often needed to satisfactorily achieve these goals. Among them, gadolinia are now largely used, owing to their excellent performance. More than 24 Framatome first cores and reloads, representing more than 3000 gadolinia-bearing rods, have been irradiated since 1983

  19. Jam Session reloaded: Von der Marmeladenfabrik zum Kultur- und Kreativraum : Revitalisierung und Umnutzung der Zuegg-Marmeladenfabrik am Tribusplatz in Lana, Südtirol

    OpenAIRE

    Hillebrand, Annika

    2015-01-01

    von Annika Hillebrand Zusammenfassung in englischer Sprache Parallelt. [Übers. des Autors]: Jam Session Reloaded: From the jam factory to the cultural and creative space revitalization and redevelopment of Zuegg jam factory on Tribusplatz in Lana, South Tyrol Technische Universität Wien, Univ., Diplomarbeit, 2015

  20. Impact of plant transient response on fuel management strategy at Virginia Power

    International Nuclear Information System (INIS)

    Bucheit, D.M.; Smith, N.A.

    1987-01-01

    Virginia Power has been performing in-house reload core design and safety analysis for several years. These analyses have been in support of North Anna units 1 and 2 and Surry units 1 and 2, all of which are three-loop pressurized water reactor plants designed and built by Westinghouse. Historically, Virginia Power first developed the capability to design and optimize its own core loading patterns in the early 1970's. This development effort was driven by the need to establish in-house control of the fuel management process, thereby ensuring that energy generation requirements are met in an economically optimum fashion. It soon became obvious that reload design and safety analysis processes are so integrally coupled that in order to perform the fuel management function in an effective manner, in-house capability in both areas needed to be developed. After reviewing the spectrum of economic, safety and operational constraints which affect the reload design and analysis process, an integrated model of the process is presented in flow chart format. This is followed by several specific examples which illustrate the interplay between sound fuel management practice and the assurance of plant safety using in-house analysis techniques

  1. Cleaning device for fuel assemblies

    International Nuclear Information System (INIS)

    Kita, Kaoru.

    1986-01-01

    Purpose: To completely remove obstacles deposited to the lower sides of supporting lattices for fuel assemblies by utilizing water within a pit before reloading of the fuel assemblies. Constitution: A cylindrical can, to which a fuel assembly is inserted through the upper end opening thereof, is vertically disposed within water of a pit and the bottom of the can is communicated with a pump by way of a suction pipe and a filter device disposed out of the pit. While on the other hand, a fuel assembly is suspended downwardly by a crane and inserted to the inside of the can through the upper end of the opening thereof and supported therein followed by starting the pump. As a result, water in the pit is circulated through the inside of the can, suction pipe, filtering device, pump, discharge pipe and to the inside of the pit thereby enabling to completely eliminate obstacles deposited to the lower surface, etc. of supporting lattices for the fuel assembly supported within the can. (Takahashi, M.)

  2. Research report on development of spacer grid strap for AFA 3G fuel assembly

    International Nuclear Information System (INIS)

    Ye Yuandong

    2004-11-01

    The current development and tendency for fuel assemblies being of low leakage, high burn-up and long cycle fuel reload in the world are presented, and the necessity and feasibility to develop the spacer grid for high burn-up fuel assembly are elaborated. Considering all the activities in implementing of spacer grid and the technical difficulties in machining of tools, the major technological processes are introduced; the research program and the approaches to develop the spacer grid while research targets and overall schedule are defined and some key technical points and applicable practices are discussed. Finally the requirements and the conditions necessary for developing of spacer grid are proposed. (authors)

  3. Fuel lattice design using heuristics and new strategies

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Pelta, D. A. [ETS Ingenieria Informatica y Telecomunicaciones, Universidad de Granada, Daniel Saucedo Aranda s/n, 18071 Granada (Spain); Campos S, Y., E-mail: juanjose.ortiz@inin.gob.m [IPN, Escuela Superior de Fisica y Matematicas, Unidad Profesional Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2010-10-15

    This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

  4. Fuel lattice design using heuristics and new strategies

    International Nuclear Information System (INIS)

    Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R.; Pelta, D. A.; Campos S, Y.

    2010-10-01

    This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

  5. Some aspects of nuclear fuel use at Ukrainian NPPs during last two years

    International Nuclear Information System (INIS)

    Bilodid, Y.; Shevchenko, I.; Ieremenko, M.; Ovdiienko, I.

    2015-01-01

    For many years SSTC NRS actively participates in licensing of fuel reloading and in the implementation of new nuclear fuel types at the nuclear power plants in Ukraine. Results of the nuclear fuel use for last years are presented in the paper. The results are based on NPP documentation submitted for licensing to the regulating body of Ukraine and based on our estimations and independent calculations. The first part of the paper contains a brief characteristic of the fuel cycles at Ukrainian NPPs. Types of loaded fuel are described also. Experience of new fuel type implementation is presented (Westinghouse FA and TVSA-12 for WWER-1000 reactors). The next part of the paper presents a new regulatory document under development and further new fuel implementation (WWER-1000 reactors). The last part of the paper describes some issues with fuel use. (authors) Keywords: WWER, TVSA, TVSA-12, TVS-W, TVS-WR, Westinghouse, NPP

  6. Fuel Management in Candu Reactors Using Tabu Search

    International Nuclear Information System (INIS)

    Chambon, R.; Varin, E.

    2008-01-01

    Meta-heuristic methods are perfectly suited to solve fuel management optimization problem in LWR. Indeed, they are originally designed for combinatorial or integer parameter problems which can represent the reloading pattern of the assemblies. For the Candu reactors the problem is however completely different. Indeed, this type of reactor is refueled online. Thus, for their design at fuel reloading equilibrium, the parameter to optimize is the average exit burnup of each fuel channel (which is related to the frequency at which each channel has to be reloaded). It is then a continuous variable that we have to deal with. Originally, this problem was solved using gradient methods. However, their major drawback is the potential local optimum into which they can be trapped. This makes the meta-heuristic methods interesting. In this paper, we have successfully implemented the Tabu Search (TS) method in the reactor diffusion code DONJON. The case of an ACR-700 using 7 burnup zones has been tested. The results have been compared to those we obtained previously with gradient methods. Both methods give equivalent results. This validates them both. The TS has however a major drawback concerning the computation time. A problem with the enrichment as an additional parameter has been tested. In this case, the feasible domain is very narrow, and the optimization process has encountered limitations. Actually, the TS method may not be suitable to find the exact solution of the fuel management problem, but it may be used in a hybrid method such as a TS to find the global optimum region coupled with a gradient method to converge faster on the exact solution. (authors)

  7. Performance and reliability of LWR fuel

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.

    1977-01-01

    The main requirements for fuel reloads are: good reliability, minimum fuel cycle costs and flexibility of operation. Fulfilling these goals requires a background of experience. The approach to the acquisition of this experience in the particular case of BN has included over the last 15 years a proper development and cross-checking of the design methods and criteria, a continuous updating of the drawings and specifications and the qualification of adequate fabrication plants. This approach can best be outlined on the basis of the gradual implementation of the modern features of the LWR fuel. The first fuel clad with stainless steel was loaded in the BR 3 (11 MWe) in 1969 and later on (since 1974) in the SENA plant (310 MWe). Similarly, Zircaloy 4 cladding was first introduced in a reactor reload in 1969 as autoclaved cladding and later on (in 1971) the autoclaving was suppressed for the further reloads. Zircaloy 2 was loaded in DODEWAARD (51.5 MWe) in 1970. The first demonstration assembly in a PWR was a Pu-island assembly loaded in the BR 3 in 1963. It was followed by an all-Pu assembly in the same reactor in 1965 and by the loading of Pu fuels in four prototype assemblies in GARIGLIANO (160 MWe) in 1968. A full reload incorporating Pu fuel has been experienced by the supply of fuel for GARIGLIANO (BOL: 1975) and for BR 3 (BOL: 1972 and 1976). While in the early sixties the brazed design was still being utilized, the first assembly incorporating grids with springs was introduced in BR 3 in 1963. The first Inconel grids were loaded in the same reactor in 1969 and the first Zircaloy grids in 1972 (the first Zr grid has been loaded in a BWR in 1973). The experience covered successively the shrouded design (BOL: 1963), the shroudless design (BOL: 1969), a BWR assembly (BOL: 1971), a typical RCC assembly first with large diameter fuel rods (1972) and later on with small diameter fuel rods (1974). The experience on the reactivity control covered successively diluted

  8. Effect of unloading followed by reloading on expression of collagen and related growth factors in rat tendon and muscle

    DEFF Research Database (Denmark)

    Heinemeier, K M; Olesen, J L; Haddad, F

    2009-01-01

    Tendon tissue and the extracellular matrix of skeletal muscle respond to mechanical loading by increased collagen expression and synthesis. This response is likely a secondary effect of a mechanically induced expression of growth factors, including transforming growth factor-beta1 (TGF-beta1......) and insulin-like growth factor-I (IGF-I). It is not known whether unloading of tendon tissue can reduce the expression of collagen and collagen-inducing growth factors. Furthermore, the coordinated response of tendon and muscle tissue to disuse, followed by reloading, is unclear. Female Sprague-Dawley rats...... tissue growth factor (CTGF), myostatin, and IGF-I isoforms were measured by real-time RT-PCR in Achilles tendon and soleus muscle. The tendon mass was unchanged, while the muscle mass was reduced by 50% after HS (P

  9. Application of CASMO-4/MICROBURN-B2 methodology to mixed cores with Westinghouse Optima2 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hsiao, Ming Yuan; Wheeler, John K.; Hoz, Carlos de la [Nuclear Fuels, Warrenville (United States)

    2008-10-15

    The first application of CASMO-4/MICROBURN-B2 methodology to Westinghouse SVEA-96 Optima2 reload cycle is described in this paper. The first Westinghouse Optima2 reload cycle in the U.S. is Exelon's Quad Cities Unit 2 Cycle 19 (Q2C19). The core contains fresh Optima2 fuel and once burned and twice burned GE14 fuel. Although the licensing analyses for the reload cycle are performed by Westinghouse with Westinghouse methodology, the core is monitored with AREVA's POWERPLEX-III core monitoring system that is based on the CASMO-4/MICROBURN-B2 (C4/B2) methodology. This necessitates the development of a core model based on the C4/B2 methodology for both reload design and operational support purposes. In addition, as expected, there are many differences between the two vendors' methodologies; they differ not only in modeling some of the physical details of the Optima2 bundles but also in the modeling capability of the computer codes. In order to have high confidence that the online core monitoring results during the cycle startup and operation will comply with the Technical Specifications requirements (e.g., thermal limits, shutdown margins), the reload core design generated by Westinghouse design methodology was confirmed by the C4/B2 model. The C4/B2 model also assures that timely operational support during the cycle can be provided. Since this is the first application of C4/B2 methodology to an Optima2 reload in the US, many issues in the lattice design, bundle design, and reload core design phases were encountered. Many modeling issues have to be considered in order to develop a successful C4/B2 core model for the Optima2/GE14 mixed core. Some of the modeling details and concerns and their resolutions are described. The Q2C19 design was successfully completed and the 2 year cycle successfully started up in April 2006 and shut down in March 2008. Some of the operating results are also presented.

  10. Application of CASMO-4/MICROBURN-B2 methodology to mixed cores with Westinghouse Optima2 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hsiao, Ming Yuan; Wheeler, John K.; Hoz, Carlos de la [Nuclear Fuels, Warrenville (United States)

    2008-10-15

    The first application of CASMO-4/MICROBURN-B2 methodology to Westinghouse SVEA-96 Optima2 reload cycle is described in this paper. The first Westinghouse Optima2 reload cycle in the U.S. is Exelon's Quad Cities Unit 2 Cycle 19 (Q2C19). The core contains fresh Optima2 fuel and once burned and twice burned GE14 fuel. Although the licensing analyses for the reload cycle are performed by Westinghouse with Westinghouse methodology, the core is monitored with AREVA's POWERPLEX-III core monitoring system that is based on the CASMO-4/MICROBURN-B2 (C4/B2) methodology. This necessitates the development of a core model based on the C4/B2 methodology for both reload design and operational support purposes. In addition, as expected, there are many differences between the two vendors' methodologies; they differ not only in modeling some of the physical details of the Optima2 bundles but also in the modeling capability of the computer codes. In order to have high confidence that the online core monitoring results during the cycle startup and operation will comply with the Technical Specifications requirements (e.g., thermal limits, shutdown margins), the reload core design generated by Westinghouse design methodology was confirmed by the C4/B2 model. The C4/B2 model also assures that timely operational support during the cycle can be provided. Since this is the first application of C4/B2 methodology to an Optima2 reload in the US, many issues in the lattice design, bundle design, and reload core design phases were encountered. Many modeling issues have to be considered in order to develop a successful C4/B2 core model for the Optima2/GE14 mixed core. Some of the modeling details and concerns and their resolutions are described. The Q2C19 design was successfully completed and the 2 year cycle successfully started up in April 2006 and shut down in March 2008. Some of the operating results are also presented.

  11. Application of CASMO-4/MICROBURN-B2 methodology to mixed cores with Westinghouse Optima2 fuel

    International Nuclear Information System (INIS)

    Hsiao, Ming Yuan; Wheeler, John K.; Hoz, Carlos de la

    2008-01-01

    The first application of CASMO-4/MICROBURN-B2 methodology to Westinghouse SVEA-96 Optima2 reload cycle is described in this paper. The first Westinghouse Optima2 reload cycle in the U.S. is Exelon's Quad Cities Unit 2 Cycle 19 (Q2C19). The core contains fresh Optima2 fuel and once burned and twice burned GE14 fuel. Although the licensing analyses for the reload cycle are performed by Westinghouse with Westinghouse methodology, the core is monitored with AREVA's POWERPLEX-III core monitoring system that is based on the CASMO-4/MICROBURN-B2 (C4/B2) methodology. This necessitates the development of a core model based on the C4/B2 methodology for both reload design and operational support purposes. In addition, as expected, there are many differences between the two vendors' methodologies; they differ not only in modeling some of the physical details of the Optima2 bundles but also in the modeling capability of the computer codes. In order to have high confidence that the online core monitoring results during the cycle startup and operation will comply with the Technical Specifications requirements (e.g., thermal limits, shutdown margins), the reload core design generated by Westinghouse design methodology was confirmed by the C4/B2 model. The C4/B2 model also assures that timely operational support during the cycle can be provided. Since this is the first application of C4/B2 methodology to an Optima2 reload in the US, many issues in the lattice design, bundle design, and reload core design phases were encountered. Many modeling issues have to be considered in order to develop a successful C4/B2 core model for the Optima2/GE14 mixed core. Some of the modeling details and concerns and their resolutions are described. The Q2C19 design was successfully completed and the 2 year cycle successfully started up in April 2006 and shut down in March 2008. Some of the operating results are also presented

  12. Time-Course of Muscle Mass Loss, Damage, and Proteolysis in Gastrocnemius following Unloading and Reloading: Implications in Chronic Diseases

    Science.gov (United States)

    Chacon-Cabrera, Alba; Lund-Palau, Helena; Gea, Joaquim; Barreiro, Esther

    2016-01-01

    Background Disuse muscle atrophy is a major comorbidity in patients with chronic diseases including cancer. We sought to explore the kinetics of molecular mechanisms shown to be involved in muscle mass loss throughout time in a mouse model of disuse muscle atrophy and recovery following immobilization. Methods Body and muscle weights, grip strength, muscle phenotype (fiber type composition and morphometry and muscle structural alterations), proteolysis, contractile proteins, systemic troponin I, and mitochondrial content were assessed in gastrocnemius of mice exposed to periods (1, 2, 3, 7, 15 and 30 days) of non-invasive hindlimb immobilization (plastic splint, I cohorts) and in those exposed to reloading for different time-points (1, 3, 7, 15, and 30 days, R cohorts) following a seven-day period of immobilization. Groups of control animals were also used. Results Compared to non-exposed controls, muscle weight, limb strength, slow- and fast-twitch cross-sectional areas, mtDNA/nDNA, and myosin content were decreased in mice of I cohorts, whereas tyrosine release, ubiquitin-proteasome activity, muscle injury and systemic troponin I levels were increased. Gastrocnemius reloading following splint removal improved muscle mass loss, strength, fiber atrophy, injury, myosin content, and mtDNA/nDNA, while reducing ubiquitin-proteasome activity and proteolysis. Conclusions A consistent program of molecular and cellular events leading to reduced gastrocnemius muscle mass and mitochondrial content and reduced strength, enhanced proteolysis, and injury, was seen in this non-invasive mouse model of disuse muscle atrophy. Unloading of the muscle following removal of the splint significantly improved the alterations seen during unloading, characterized by a specific kinetic profile of molecular events involved in muscle regeneration. These findings have implications in patients with chronic diseases including cancer in whom physical activity may be severely compromised. PMID

  13. Influence of the faces relative arrangement on the optimal reloading station location and analytical determination of its coordinates

    Directory of Open Access Journals (Sweden)

    V.К. Slobodyanyuk

    2017-04-01

    Full Text Available The purpose of this study is to develop a methodology of the optimal rock mass run-of-mine (RoM stock point determination and research of the influence of faces spatial arrangement on this point. The research represents an overview of current researches, where algorithms of the Fermat-Torricelli-Steiner point are used in order to minimize the logistic processes. The methods of mathematical optimization and analytical geometry were applied. Formulae for the optimal point coordinates determination for a 4 faces were established using the latter methods. Mining technology with use of reloading stations is rather common at the deep iron ore pits. In most cases, when deciding on location of RoM stock, its high-altitude position in space of the pit is primarily taken into account. However, the location of the reloading station in a layout also has a significant influence on technical and economic parameters of open-pit mining operations. The traditional approach, which considers a point of the center of gravity as an optimal point for RoM stock location, does not guarantee the minimum haulage. In mathematics, the Fermat-Torricelli point that provides a minimum distance to the vertices of the triangle is known. It is shown that the minimum haulage is provided when the point of RoM stock location and Fermat-Torricelli point coincide. In terms of open pit mining operations, the development of a method that will determine an optimal point of RoM stock location for a working area with respect to the known coordinates of distinguished points on the basis of new weight factors is of particular practical importance. A two-stage solution to the problem of determining the rational point of RoM stock location (with a minimal transport work for any number of faces is proposed. Such optimal point for RoM stock location reduces the transport work by 10–20 %.

  14. System to solve three designs of the fuel management; Sistema para resolver tres disenos de la administracion de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Castillo M, J. A.; Ortiz S, J. J.; Montes T, J. L.; Perusquia del C, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Marinez R, R., E-mail: alejandro.castillo@inin.gob.mx [Universidad Autonoma de Campeche, Av. Agustin Melgar s/n, Col. Buenavista, 24039 San Francisco de Campeche, Campeche (Mexico)

    2015-09-15

    In this paper preliminary results are presented, obtained with the development of a computer system that resolves three stages of the nuclear fuel management, which are: the axial and radial designs of fuel, as well as the design of nuclear fuel reloads. The novelty of the system is that the solution is obtained solving the 3 mentioned stages, in coupled form. For this, heuristic techniques are used for each stage, in each one of these has a function objective that is applied to particular problems, but in all cases the obtained partial results are used as input data for the next stage. The heuristic techniques that were used to solve the coupled problem are: tabu search, neural networks and a hybrid between the scatter search and path re linking. The system applies an iterative process from the design of a fuel cell to the reload design, since are preliminary results the reload is designed using the operation strategy Haling type. In each one of the stages nuclear parameters inherent to the design are monitored. The results so far show the advantage of solving the problem in a coupled manner, even when a large amount of computer resources is used. (Author)

  15. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)

    2015-03-15

    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  16. Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

    International Nuclear Information System (INIS)

    Hida, Kazuki; Yoshioka, Ritsuo

    1992-01-01

    A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

  17. Coupling of the four design stages in the management of nuclear fuel

    International Nuclear Information System (INIS)

    Marinez R, R. L.

    2016-01-01

    In this work, the main characteristics of the system to solve the four stages of the nuclear fuel management are presented; the above for boiling water reactors (BWR). The novelty of the system is that a complete solution is obtained in a coupled way; the involved stages are fuel lattice design, fuel assembly design, fuel reload design and control rod pattern design. To do this, in each stage of the process some heuristics techniques are applied, and each stage has its own objective function. The used heuristic techniques are neural network and a hybrid between scatter search and path re linking for fuel lattice design; for fuel assembly design a simple local search was applied and finally, for both fuel reload and control rod pattern designs, the tabu search technique was used. The system have two loops, one external loop and one internal loop, the first one starts with fuel lattice design and concludes with control rod pattern design; on the other hand, the internal loop executes an iterative process between both fuel reload design and control rod pattern designs, to start this loop a seed fuel reload is required, which is obtained applying Haling principle. The internal loop is finished when four iterations were achieved, while the external loop is finished when two iterations were achieved, this number of iterations was fixed due to the great quantity of required computational resources. An 18- months equilibrium cycle was considered to have a reference value to compare against the obtained results with our system, this cycle have two fuel fresh batches with the same average uranium enrichment, but different gadolinia content. The above cycle achieved a 10,896 Mwd/Tu of energy and was divided into 12 burnup steps. The obtained results show the advantage to solve the complete problem in a coupled way, even though a great quantity of computational resources are used. It is necessary to note that the energy value was not achieved in all cases, only in some

  18. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  19. Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

    International Nuclear Information System (INIS)

    Roudier, S.; Badel, D.; Beraha, R.; Champ, M.; Tricot, N.; Tran Dai, P.

    1994-01-01

    The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: 1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; 2) guidelines for nuclear design and manufacturing requirements related to safety and 3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs

  20. Practice and trends in nuclear fuel licensing in France (pressurized water reactor fuels)

    Energy Technology Data Exchange (ETDEWEB)

    Roudier, S [Direction de la Surete des Installations Nucleaires, Fontenay-aux-Roses (France); Badel, D; Beraha, R [Direction Regionale de l` Industrie, de la Recherche et de l` Environnement Rhone-Alpes, Lyon (France); Champ, M; Tricot, N; Tran Dai, P [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire

    1994-12-31

    The activities of governmental French authorities responsible for safety of nuclear installations are outlined. The main bodies involved in nuclear safety are: the CSSIN (High Council for Nuclear Safety and Information), CINB (Inter-ministerial Commission for Basic Nuclear Installations) and DSIN (Nuclear Installations Safety Directorate). A brief review of the main fuel licensing issues supported by DSIN is given, which includes: (1) formal regularity procedure ensuring the safety of nuclear installations and especially the pressurized water reactors; (2) guidelines for nuclear design and manufacturing requirements related to safety and (3) safety goals and associated limits. The fuel safety documents for reloading as well as the research and development programmes in the field of technical safety are also described. The ongoing experiments in CABRI reactor, aimed at determining the high burnup fuel behaviour under reactivity initiated accidents until 65 GW d/Mt U, are one of these programs.

  1. Nuclear Fuel Cycle Evaluation and Real Options

    Directory of Open Access Journals (Sweden)

    L. Havlíček

    2008-01-01

    Full Text Available The first part of this paper describes the nuclear fuel cycle. It is divided into three parts. The first part, called Front-End, covers all activities connected with fuel procurement and fabrication. The middle part of the cycle includes fuel reload design activities and the operation of the fuel in the reactor. Back-End comprises all activities ensuring safe separation of spent fuel and radioactive waste from the environment. The individual stages of the fuel cycle are strongly interrelated. Overall economic optimization is very difficult. Generally, NPV is used for an economic evaluation in the nuclear fuel cycle. However the high volatility of uranium prices in the Front-End, and the large uncertainty of both economic and technical parameters in the Back-End, make the use of NPV difficult. The real option method is able to evaluate the value added by flexibility of decision making by a company under conditions of uncertainty. The possibility of applying this method to the nuclear fuel cycle evaluation is studied. 

  2. Worldwide supply of Framatome ANP Fuel

    International Nuclear Information System (INIS)

    Jouan, J.

    2002-01-01

    Framatome-ANP is organized according to a matrix structure with 4 business groups and 3 regional divisions. The fuel business group with a workforce of about 4600 people is active in all the trades needed to design and manufacture nuclear fuel. The activity ranges from the production of zirconium alloys to the production of finished fuel assemblies, facilities are located in France, Germany and Usa. Framatome-ANP is the foremost vendor of LWR fuel worldwide with 41 % of the PWR market share and 22 % of the BWR market share. The global operating experience built up is based on more than 150.000 fuel assemblies delivered to 169 reactors in 18 countries. This long history has allowed Framatome-ANP to develop an efficient quality-improvement program based on experience feedback, for instance fuel rod failures induced by debris have been almost completely eliminated with the introduction of anti-debris devices equipping bottom nozzles. Framatome-ANP has developed a large range of engineering services, for instance core design teams can provide the most cost-effective fuel management schemes for cycle lengths from 6 to 24 months. The first technology transfer between China entities and Framatome related to the AFA-2G technology started in 1991 and was completed successfully in 1994. Since this date the Chinese manufacturer has supplied fuel reload for the units of Daya-Bay. (A.C.)

  3. 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

  4. Fuel deposits and water chemistry at TVO I power station during the first three fuel cycles

    International Nuclear Information System (INIS)

    Silvennoinen, S.; Hakala, J.

    1983-01-01

    TVO 1 is a 660 MWe direct-cycle, light-water cooled BWR of Asea-Atom design. The unit has just completed the 4th cycle. Sampling of deposits on fuel surfaces has been performed by Asea-Atom after each cycle. The deposits consist mainly of iron (78-86%) excepting two rods in a bundle belonging to the first reload. On these two rods the amounts of Cr and Cu were exceptionally high indicating differences in the deposition processes between the initial fuel and the first reload. In general the crud thickness is highest at a height of 1 to 1.5 meters from the bottom plate of the bundle. The average concentrations of the corrosion products vary from 3.6 to 10.3 g/m 2 /bundle. Inexplicable large variations of the crud amount between the bundles and between the individual rods in a bundle have been measured. Growth rate of deposits is decreased and water chemistry is improved with improved operation of the plant. During normal operation many impurities are below the detection limit thus making the interpretation of the results difficult. (author)

  5. 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

  6. Conversion of highly enriched uranium in thorium-232 based oxide fuel for light water reactors: MOX-T fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vapirev, E I; Jordanov, T; Christoskov, I [Sofia Univ. (Bulgaria). Fizicheski Fakultet

    1994-12-31

    The idea of conversion of highly enriched uranium (HEU) from warheads without mixing it with natural uranium as well as the utilization of plutonium as fuel component is discussed. A nuclear fuel which is a mixture of 4% {sup 235}U (HEU) as a fissile isotope and 96 % {sup 232}Th (ThO{sub 2}) as a non-fissile isotope in a mixed oxide with thorium fuel is proposed. It is assumed that plutonium can also be used in the proposed fuel in a mixture with {sup 235}U. The following advantages of the use of HEU in LWRs in mixed {sup 235}U - Th fuel are pointed out: (1) No generation of long-living plutonium and americium isotopes (in case of reprocessing the high level radioactive wastes will contain only fission fragments and uranium); (2) The high conversion ratio of Th extends the expected burnup by approximately 1/3 without higher initial enrichment (the same initial enrichment simplifies the problem for compensation of the excess reactivity in the beginning with burnable poison and boric acid); (3) The high conversion ratio of Th allows the fuel utilization with less initial enrichment (by approx. 1/3) for the same burnup; thus less excess reactivity has to be compensated after reloading; in case of fuel reprocessing all fissile materials ({sup 235}U + {sup 233}U) could be chemically extracted. Irrespectively to the optimistic expectations outlined, further work including data on optimal loading and reloading schemes, theoretical calculations of thermal properties of {sup 235}U + Th fuel rods, manufacturing of several test fuel assemblies and investigations of their operational behaviour in a reactor core is still needed. 1 fig., 7 refs.

  7. General considerations in fuel management for thermal reactors

    International Nuclear Information System (INIS)

    Tyror, J.G.; Fayers, F.J.

    1971-07-01

    By fuel management we mean the strategy for fuelling and refuelling a reactor together with any associated absorber movements. It incorporates (a) decisions made about the timing of fuel loading operations; (b) choice of enrichments to be loaded; (c) selection of sites at which reloading occurs; (d) programming of control rods and any other reactivity control facilities such as soluble or burnable poisons; and (e) evaluation of the resulting fuel element performance consequences. The topic of fuel management is thus a vast and vital one. It embraces most of the various aspects of core performance and determines many of a reactor's design characteristics. In this paper we review what to us appear to be some of the important issues in this important field

  8. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  9. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    International Nuclear Information System (INIS)

    2009-06-01

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

  10. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Fourth semiannual report, July-December 1980

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1981-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts have been developed for possible demonstration: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the scope of this program one of these concepts had to be selected for a large-scale demonstration in a commercial power reactor. The selection was made to demonstrate Zr-liner fuel and to include bundles which have liners prepared from either low oxygen sponge zirconium or of crystal bar zirconium. The demonstration is intended to include a total of 132 barrier bundles in the reload for Quad Cities Unit 2, Cycle 6. In the current report period changes in the nuclear design were made to respond to changes in the Energy Utilization Plan for Quad Cities Unit 2. Bundle designs were completed, and were licensed for use in a BWR/3. The core specific licensing will be done as part of the reload license for Quad Cities Unit 2, Cycle 6

  11. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    International Nuclear Information System (INIS)

    Beard, Ch.; Morita, T.; Brown, J.

    2007-01-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  12. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    Energy Technology Data Exchange (ETDEWEB)

    Beard, Ch.; Morita, T.; Brown, J. [Westinghouse Electric Company, LLC, Nuclear Fuel Div., Pittsburgh, PA (United States)

    2007-07-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  13. Fuel management optimization based on generalized perturbation theory

    International Nuclear Information System (INIS)

    White, J.R.; Chapman, D.M.; Biswas, D.

    1986-01-01

    A general methodology for optimization of assembly shuffling and burnable poison (BP) loadings for LWR reload design has been developed. The uniqueness of this approach lies in the coupling of Generalized Perturbation Theory (GPT) methods and standard Integer Programming (IP) techniques. An IP algorithm can simulate the discrete nature of the fuel shuffling and BP loading problems, and the use of GPT sensitivity data provides an efficient means for modeling the behavior of the important core performance parameters. The method is extremely flexible since the choice of objective function and the number and mix of constraints depend only on the ability of GPT to determine the appropriate sensitivity functions

  14. 8 x 8 fuel surveillance program at Monticello site - end of Cycle 6: fourth post-irradiation inspection, October 1978

    International Nuclear Information System (INIS)

    Skarshaug, N.H.

    1980-09-01

    A fuel surveillance program for a lead 8 x 8 reload fuel assembly was implemented at the Monticello Nuclear Power Station in May 1974 prior to Reactor Cycle 3. Inspection results of the fourth post-irradiation inspection performed on this surveillance fuel assembly in October 1978 at EOC 6, after a bundle average exposure of 25,900 MWd/MT, are presented. The measurement techniques, results obtained and comparisons to previous measurements are discussed. The bundle and individual rods examined exhibited characteristics of normal operation and were approved for continued irradiation during Monticello operating Cycle 7

  15. Fuel management for off-load annual refuelling of the D-HHT 600 MW(e) reference core

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1973-03-16

    The reference design for the Dragon-HHT reactor has been optimised for on-load continuous refuelling. The possiblity to operate the reactor on a discontinuous annual reloading schedule might prove of interest and/or necessity. In this paper the influence of an annual 4-batch fuel management scheme on the core physics and fuel cycle economics is investigated. The results of the present investigation give a good indication of the relative merits of the two fuel management schemes. Although a broader parameter survey and a more detailed scrutinising of special cases would be desirable, we feel that the main conclusions are correct and that the principle differences have been elicited.

  16. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  17. Fuel management options to extend the IRIS reactor cycle

    International Nuclear Information System (INIS)

    Petrovic, B.; Franceschini, F.

    2004-01-01

    To optimize plant operation, reduce scheduled maintenance outage, and increase capacity factor, IRIS is designed to enable extended cycles of up to four years. However, due to the enrichment licensing limitation (less than 5% enriched uranium oxide) there is a tradeoff between the achievable cycle length and fuel utilization, i.e., the average fuel discharge burnup. The longest individual cycle may be achieved with the single-batch straight burn, but at the expense of a lower burnup. Considering the IRIS basic performance requirements, a cycle length in the range of three to four years is deemed desirable. This paper examines different fuel management options, i.e., the influence of the required cycle length on the corresponding reloading strategy, including a two-batch and a three-batch reloading. A reference two-batch core design has been developed for the first cycle, as well as for the transition cycles leading to equilibrium. Main core performance parameters are evaluated. This core design provides the framework for the safety analyses needed to prepare the IRIS safety evaluations. Alternate designs are also considered.(author)

  18. Operational experience for the latest generation of ATRIUM trademark 10 fuel assemblies

    International Nuclear Information System (INIS)

    Schoss, Volker; Hoffmann, Petra Britt; Schaefer, Jens

    2011-01-01

    AREVA NP's ATRIUM trademark 10 product family was first introduced to the BWR market in 1992. Lead test campaigns confirmed the outstanding product performance and justified introduction of reload quantities. Further development of particular product features was demonstrated and implemented in the fuel design to meet highest expectations for reliability and fuel economics. The latest generation called ATRIUM trademark 10XP and subsequently ATRIUM trademark 10XM was introduced in 2002 and 2005, respectively. The first lead test assemblies completed their operation successfully after seven cycles. (orig.)

  19. A calculation methodology applied for fuel management in PWR type reactors using first order perturbation theory

    International Nuclear Information System (INIS)

    Rossini, M.R.

    1992-01-01

    An attempt has been made to obtain a strategy coherent with the available instruments and that could be implemented with future developments. A calculation methodology was developed for fuel reload in PWR reactors, which evolves cell calculation with the HAMMER-TECHNION code and neutronics calculation with the CITATION code.The management strategy adopted consists of fuel element position changing at the beginning of each reactor cycle in order to decrease the radial peak factor. The bi-dimensional, two group First Order perturbation theory was used for the mathematical modeling. (L.C.J.A.)

  20. Analysis and application of advanced fuel management strategies at Florida Power and Light Company

    International Nuclear Information System (INIS)

    Knuckles, E.R.; Mantyh, J.D.; Hoskins, K.C.

    1986-01-01

    Reload design flexibility is the degree of freedom that the fuel management engineer has in utilizing various options to achieve and maintain an optimal core design. The major factors affecting flexibility are the basic design constraints, fuel design, and operational and regulatory uncertainty. The degree of flexibility available to the engineer can be improved through an understanding of the inter-relationship of these factors. Specific examples are used to demonstrate how the concept of flexibility has been implemented at Florida Power and Light Company

  1. REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR

    International Nuclear Information System (INIS)

    Boettcher, W.; Schmidt, E.

    1969-01-01

    1 - Nature of physical problem solved: REFLOS is a programme for the evaluation of fuel-loading schemes in heavy water moderated reactors. The problems involved in this study are: a) Burn-up calculation for the reactor cell. b) Determination of reactivity behaviour, power distribution, attainable burn-up for both the running-in period and the equilibrium of a 3-dimensional heterogeneous reactor model; investigation of radial fuel movement schemes. c) Evaluation of mass flows of heavy atoms through the reactor and fuel cycle costs for the running-in, the equilibrium, and the shut down of a power reactor. If the subroutine for treating the reactor cell were replaced by a suitable routine, other reactors with weakly absorbing moderators could be analyzed. 2 - Method of solution: Nuclear constants and isotopic compositions of the different fuels in the reactor are calculated by the cell-burn-up programme and tabulated as functions of the burn-up rate (MWD/T). Starting from a known state of the reactor, the 3-dimensional heterogeneous reactor programme (applying an extension of the technique of Feinberg and Galanin) calculates reactivity and neutron flux distribution using one thermal and one or two fast neutron groups. After a given irradiation time, the new state of the reactor is determined, and new nuclear constants are assigned to the various defined locations in the reactor. Reloading of fuel may occur if the prescribed life of the reactor is reached or if the effective multiplication factor or the power form factor falls below a specified level. The scheme of reloading to be carried out is specified by a load vector, giving the number of channels to be discharged, the kind of movement from one to another channel and the type of fresh fuel to be charged for each single reloading event. After having determined the core states characterizing the equilibrium period, and having decided the fuel reloading scheme for the running-in period of the reactor life, the fuel

  2. LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

    International Nuclear Information System (INIS)

    Peehs, M.

    1997-01-01

    Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves

  3. LWR FA burn up: A challenge to optimize the entire fuel cycle to assure the envisaged benefit

    Energy Technology Data Exchange (ETDEWEB)

    Peehs, M [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-12-01

    Commercial LWR fuel will be limited to a maximum of U-235 content of 5% since the front end of the fuel cycle is licensed and prepared for that maximal enrichment. BWR- and PWR-reloads can be designed achieving batch average burn up over 60 GWd/tHM. In Germany the batch average burn up will presumably increase to this level, since the reload market is requesting further reductions in the fuel cycle inventories. However, it must be noted that the envisaged benefit can only be assured if the entire fuel cycle is optimized. Not all steps in the fuel cycle will bring a positive contribution bu the balance of all individual contributions must realize the envisaged integral benefit. In order to increase the burn up of the nuclear fuel beneficially further R and D both in the front end as well as in the back end of the fuel cycle is needed. An underestimation of the front end/back end interfaces may consume all benefits gained from isolated front optimizations. Back end R and D must be at once concentrated to avoid conservative enveloping licensing for the subsequent steps in the back end of the fuel cycle. Increasing burn up in the front end means making more and more use of the structural materials reserves.

  4. Discussion on the re-irradiated fuel assembly with damaged guide vanes

    International Nuclear Information System (INIS)

    Li Ligang

    2013-01-01

    In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

  5. A feasible approach to implement a commercial scale CANDU fuel manufacturing plant in Egypt

    International Nuclear Information System (INIS)

    El-Shehawy, I.; El-Sharaky, M.; Yasso, K.; Selim, I.; Graham, N.; Newington, D.

    1995-01-01

    Many planning scenarios have been examined to assess and evaluate the economic estimates for implementing a commercial scale CANDU fuel manufacturing plant in Egypt. The cost estimates indicated strong influence of the annual capital costs on total fuel manufacturing cost; this is particularly evident in a small initial plant where the proposed design output is only sufficient to supply reload fuel for a single CANDU-6 reactor. A modular approach is investigated as a possible way, to reduce the capital costs for a small initial fuel plant. In this approach the plant would do fuel assembly operations only and the remainder of a plant would be constructed and equipped in the stages when high production volumes can justify the capital expenses. Such approach seems economically feasible for implementing a small scale CANDU fuel manufacturing plant in developing countries such as Egypt and further improvement could be achieved over the years of operation. (author)

  6. Risk-based decision making for staggered bioterrorist attacks : resource allocation and risk reduction in "reload" scenarios.

    Energy Technology Data Exchange (ETDEWEB)

    Lemaster, Michelle Nicole; Gay, David M. (Sandia National Laboratories, Albuquerque, NM); Ehlen, Mark Andrew (Sandia National Laboratories, Albuquerque, NM); Boggs, Paul T.; Ray, Jaideep

    2009-10-01

    Staggered bioterrorist attacks with aerosolized pathogens on population centers present a formidable challenge to resource allocation and response planning. The response and planning will commence immediately after the detection of the first attack and with no or little information of the second attack. In this report, we outline a method by which resource allocation may be performed. It involves probabilistic reconstruction of the bioterrorist attack from partial observations of the outbreak, followed by an optimization-under-uncertainty approach to perform resource allocations. We consider both single-site and time-staggered multi-site attacks (i.e., a reload scenario) under conditions when resources (personnel and equipment which are difficult to gather and transport) are insufficient. Both communicable (plague) and non-communicable diseases (anthrax) are addressed, and we also consider cases when the data, the time-series of people reporting with symptoms, are confounded with a reporting delay. We demonstrate how our approach develops allocations profiles that have the potential to reduce the probability of an extremely adverse outcome in exchange for a more certain, but less adverse outcome. We explore the effect of placing limits on daily allocations. Further, since our method is data-driven, the resource allocation progressively improves as more data becomes available.

  7. BWR simulation in a stationary state for the evaluation of fuel cell design

    International Nuclear Information System (INIS)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A.

    2014-10-01

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  8. Coherence of reactor design and fuel element design

    International Nuclear Information System (INIS)

    Vom Scheidt, S.

    1995-01-01

    Its background of more than 25 years of experience makes Framatome the world's leading company in the design and sales of fuel elements for pressurized water reactors (PWR). In 1994, the fuel fabrication units were incorporated as subsidiaries, which further strengthens the company's position. The activities in the fuel sector comprise fuel element design, selection and sourcing of materials, fuel element fabrication, and the services associated with nuclear fuel. Design responsibility lies with the Design and sales Management, which closely cooperates with the engineers of the reactor plant for which the fuel elements are being designed, for fuel elements are inseparable parts of the respective reactors. The Design and Sales Management also has developed a complete line of services associated with fuel element inspection and repair. As far as fuel element sales are concerned, Framatome delivers the first core in order to be able to assume full responsibility vis-a-vis the customer for the performance of the nuclear steam supply system. Reloads are sold through the Fragema Association established by Framatome and Cogema. (orig.) [de

  9. Introduction of fuel GE14 in the nuclear power plant of Laguna Verde for the extended increase of power

    International Nuclear Information System (INIS)

    Hernandez M, N.; Vargas A, A. F.; Cardenas J, J. B.; Contreras C, P.

    2008-01-01

    The project of extended increase of power responds to a necessity of electrical energy in the country, increasing the thermal exit of the reactors of the nuclear power plant of Laguna Verde of 2027 MWt to 2317 MWt. In order to support this transition, changes will make in the configuration of the reactor core and in the operation strategies of the cycle, also they will take initiatives to optimize the economy in fuel cycle. At present in both reactors of the nuclear plant of Laguna Verde fuel GE12 is used. The fuel GE14 presents displays with respect to the GE12, some improvements in the mechanical design and consequently in its performance generally. Between these improvements we can mention: 1. Spacers of high performance. 2. Shielding with barrier. 3. Filter for sweepings d ebris a nd 4. Fuel rods of minor partial length. The management of nuclear power plants has decided to introduce the use of fuel GE14 in Laguna Verde in the reload 14 for Unit 1 and of the reload 10 for Unit 2. The process of new introduction fuel GE14 consists of two stages, first consists on subjecting the one new design of fuel to the regulator organism in the USA: Nuclear Regulatory Commission, in Mexico the design must be analyzed and authorized by the National Commission of Nuclear Safety and Safeguards, for its approval of generic form, by means of the demonstration of the fulfillment with the amendment 22 of GESTAR II, the second stage includes the specific analyses of plant to justify the use of the new fuel design in a reload core. The nuclear plant of Laguna Verde would use some of the results of the security analyses that have been realized for the project of extended increase of power with fuel GE14, to document the specific analyses of plant with the new fuel design. The result of the analyses indicates that the reload lots are increased of 116-120 assemblies in present conditions (2027 MWt) to 140-148 assemblies in conditions of extended increase of power (2317 MWt). (Author)

  10. ETRR-2 in-core fuel management strategy

    International Nuclear Information System (INIS)

    Khalil, M.Y.; Amin, Esmat; Belal, M.G.

    2005-01-01

    The Egypt second research reactor has many irradiation channels, beam tubes and irradiation boxes, inside and outside the reactor core. The core reload configuration has great effect on the core performance and fluxes in the irradiation channels. This paper deals with the design and safety analysis that were performed for the determination of ETRR2 in-core fuel management strategy which fulfills neutronic design criteria, safety reactor operation, utility optimization and achieve the overall fuel management criteria. The core is divided into 8 zones, in order to obtain the minimum and adjacent fuel movement scheme that is recommended from the operational point of view. Then a search for the initial core using backward iteration, one get different initial cores, one initial core would assume the equilibrium core after 250 full power days of operation, while the other assumes equilibrium after 199 full power days, and shows a better performance of power peaking factor. (author)

  11. Where U.S. utilities seek fuel to power reactors after 1985

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    How utilities try to assure uranium supplies emerged Feb. 25 at a press conference in Canberra with four representatives of utilities that supply 20% of the operating nuclear capacity in the United States. Earlier, the speakers indicated that American import requirements would far exceed Australian estimates of the potential export market for Australian uranium. Australia, with the world's largest uncommitted uranium reserves, is wary of exporting because the opposition Labor Party adamantly opposes uranium development. If Labor returns to power, it could decide not to honor contracts by the present government. Participants included: Bernard Cherry, fuel manager at General Public Utilities; Colin Campbell of the Yankee Atomic Service Co., which provides engineering and fuel-supply service for seven New England nuclear plants; Jack Gilleland, assistant manager of power at the Tennessee Valley Authority; and Ralph Bostian, manager for systems results and fuel management at the Duke Power Co. When asked about available uranium supplies from Africa, the participants were dubious about those supplies; this led to a discussion on why the utilities are seeking their own sources. The answers are obvious. ERDA has indicated that about one-half of the operating reactors have fuel coverage beyond six reloads and about one-half of the reactors under construction have fuel coverage beyond two reloads

  12. A classification scheme for LWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

  13. A classification scheme for LWR fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs

  14. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of the nucleoelectrical generation in Mexico by 1976 is described: two nuclear reactors under construction but no defined program on the type and start-up dates for the next power plants. However the existence of a general plan on nuclear power plants is mentioned, which, according to the last estimates reaches to 10,000 MW installed by 1990. The national intension, definitely expressed in the Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reload for the two BWR's at the first national station in Laguna Verde, which will be required at the end of 1981 and of 1982, respectively. Before such circumstances and the relatively short amounts of fuel elements that should be produced for those two unique reactors, Mexico already has to adopt a strategy to follow in respect to fuel elements fabrication. The two main options are analyzed: 1. To delay the local fabrication until a National Nuclear Program may be defined, meanwhile purchasing abroad the necessary reloads and initial cores; and 2. To start as soon as possible the local fuel elements fabrication in order to supply fuel for the first reload of the first unit of Laguna Verde, confronting the economical risks of such posture with the advantages of an immediate action. Both options are analyzed in detail comparing them specially under the economic point of view, standing out immediately the big effect of some factors which are economically imponderable, as experience and independance that would be gained with the second option. Emphasis is made on the advantages and risks of any case. According to the first option and once a National Program is defined, the work would be heavy but of simple strategy. On the contrary, the second option requires the adoption of a more complicated strategy, as either the project of the factory as its initial operation should be made under transient conditions, in view of the expected future expansion still

  15. EDF advanced fuel management strategies for the next century

    International Nuclear Information System (INIS)

    Kocher, A.; Charmensat, P.; Larderet, M.

    1999-01-01

    The French nuclear fleet represents 57 PWRs in operation, accounting for 80 % of France's total electricity production. The performance achieved by EDF reactors, in terms of availability (82.6% in 1997) and good cost control, have allowed to improve the nuclear KWh cost by 2% since 1992. The implementation of longer fuel cycles on the 1300 MW reactors from 1996 has contributed to this improvement and, as competitiveness is one of the main challenges for EDF, improving core management strategies is still at the order of the day. With this aim, a thinking process has been initiated to evaluate the benefit brought by the use of a fuel assembly like ALLIANCE, the new fuel product developed by Framatome-Fragema and FCF (Framatome Cogema Fuels) in close cooperation with EDF. The considered product provides enhanced performance, particularly as regards discharge burnup (at least up to 70 GWd/t) and thermal-hydraulic and mechanical behaviour. Fuel management improvements rely on the expertise gained by Framatome through designing core management strategies in a wide range of operating conditions prevailing in nuclear reactors all over the world. It will however be taken into account the necessity for EDF to adopt a policy of stepwise change owing to the potential impact of a 'series effect' on its numerous units. The proposed paper will describe innovative fuel managements, achievable thanks to advanced fuel assembly performance, that are jointly investigated by EDF and Framatome. It includes the following optimization schemes: extending cycle length by using higher enrichments up to 5%, while keeping the same reload size (1/3 core for example for the 1300 MW reactors); decreasing reload size (from 1/3 to 1/4 core), while keeping the same cycle length, using more enriched (up to 5 %) fuel assemblies; reaching annual cycle, with maximization of fuel cycle cost optimization (1/5 core). Beyond such schemes, combinations of optimized loading patterns and neutronic features of

  16. Singlet oxygen sensitizing materials based on porous silicone: photochemical characterization, effect of dye reloading and application to water disinfection with solar reactors.

    Science.gov (United States)

    Manjón, Francisco; Santana-Magaña, Montserrat; García-Fresnadillo, David; Orellana, Guillermo

    2010-06-01

    Photogeneration of singlet molecular oxygen ((1)O(2)) is applied to organic synthesis (photooxidations), atmosphere/water treatment (disinfection), antibiofouling materials and in photodynamic therapy of cancer. In this paper, (1)O(2) photosensitizing materials containing the dyes tris(4,4'-diphenyl-2,2'-bipyridine)ruthenium(II) (1, RDB(2+)) or tris(4,7-diphenyl-1,10-phenanthroline)ruthenium(II) (2, RDP(2+)), immobilized on porous silicone (abbreviated RDB/pSil and RDP/pSil), have been produced and tested for waterborne Enterococcus faecalis inactivation using a laboratory solar simulator and a compound parabolic collector (CPC)-based solar photoreactor. In order to investigate the feasibility of its reuse, the sunlight-exposed RDP/pSil sensitizing material (RDP/pSil-a) has been reloaded with RDP(2+) (RDP/pSil-r). Surprisingly, results for bacteria inactivation with the reloaded material have demonstrated a 4-fold higher efficiency compared to those of either RDP/pSil-a, unused RDB/pSil and the original RDP/pSil. Surface and bulk photochemical characterization of the new material (RDP/pSil-r) has shown that the bactericidal efficiency enhancement is due to aggregation of the silicone-supported photosensitizer on the surface of the polymer, as evidenced by confocal fluorescence lifetime imaging microscopy (FLIM). Photogenerated (1)O(2) lifetimes in the wet sensitizer-doped silicone have been determined to be ten times longer than in water. These facts, together with the water rheology in the solar reactor and the interfacial production of the biocidal species, account for the more effective disinfection observed with the reloaded photosensitizing material. These results extend and improve the operational lifetime of photocatalytic materials for point-of-use (1)O(2)-mediated solar water disinfection.

  17. Mox fuel experience: present status and future improvements

    International Nuclear Information System (INIS)

    Blanpain, P.; Chiarelli, G.

    2001-01-01

    Up to December 2000, more than 1700 MOX fuel assemblies have been delivered by Framatome ANP/Fragema to 20 French, 2 Belgian and 3 German PWRs. More than 1000 MOX fuel assemblies have been delivered by Framatome ANP GmbH (formerly Siemens) to 11 German PWRs and BWRs and to 3 Swiss PWRs. Operating MOX fuel up to discharge burnups of about 45,000 MWd/tM is done without any penalty on core operating conditions and fuel reliability. Performance data for fuel and materials have been obtained from an outstanding surveillance program. The examinations have concluded that there have been no significant differences in MOX fuel assembly characteristics relative to UO 2 fuel. The data from these examinations, combined with a comprehensive out-of-core and in-core analytical test program on the current fuel products, are being used to confirm and upgrade the design models necessary for the continuing improvement of the MOX product. As MOX fuel has reached a sufficient maturity level, the short term step is the achievement of the parity between UO 2 and MOX fuels in the EdF French reactors. This involves a single operating scheme for both fuels with an annual quarter core reload type and an assembly discharge burnup goal of 52,000 MWd/tM. That ''MOX parity'' product will use the AFA-3G assembly structure which will increase the fuel rod design margins with regards to the end-of-life internal pressure criteria. But the fuel development objective is not limited to the parity between the current MOX and UO 2 products: that parity must remain guaranteed and the MOX fuel managements must evolve in the same way as the UO 2 ones. The goal of the MOX product development program underway in France is the demonstration over the next ten years of a fuel capable of reaching assembly burnups of 70,000 MWd/tM. (author)

  18. Coupling of the four design stages in the management of nuclear fuel; Acoplamiento de las cuatro etapas de diseno en la administracion de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Marinez R, R. L.

    2016-07-01

    In this work, the main characteristics of the system to solve the four stages of the nuclear fuel management are presented; the above for boiling water reactors (BWR). The novelty of the system is that a complete solution is obtained in a coupled way; the involved stages are fuel lattice design, fuel assembly design, fuel reload design and control rod pattern design. To do this, in each stage of the process some heuristics techniques are applied, and each stage has its own objective function. The used heuristic techniques are neural network and a hybrid between scatter search and path re linking for fuel lattice design; for fuel assembly design a simple local search was applied and finally, for both fuel reload and control rod pattern designs, the tabu search technique was used. The system have two loops, one external loop and one internal loop, the first one starts with fuel lattice design and concludes with control rod pattern design; on the other hand, the internal loop executes an iterative process between both fuel reload design and control rod pattern designs, to start this loop a seed fuel reload is required, which is obtained applying Haling principle. The internal loop is finished when four iterations were achieved, while the external loop is finished when two iterations were achieved, this number of iterations was fixed due to the great quantity of required computational resources. An 18- months equilibrium cycle was considered to have a reference value to compare against the obtained results with our system, this cycle have two fuel fresh batches with the same average uranium enrichment, but different gadolinia content. The above cycle achieved a 10,896 Mwd/Tu of energy and was divided into 12 burnup steps. The obtained results show the advantage to solve the complete problem in a coupled way, even though a great quantity of computational resources are used. It is necessary to note that the energy value was not achieved in all cases, only in some

  19. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.

    1981-01-01

    A nuclear fuel loading apparatus, incorporating a microprocessor control unit, is described which automatically loads nuclear fuel pellets into dual fuel rods with a minimum of manual involvement and in a manner and sequence to ensure quality control and accuracy. (U.K.)

  20. Sophisticated fuel handling system evolved

    International Nuclear Information System (INIS)

    Ross, D.A.

    1988-01-01

    The control systems at Sellafield fuel handling plant are described. The requirements called for built-in diagnostic features as well as the ability to handle a large sequencing application. Speed was also important; responses better than 50ms were required. The control systems are used to automate operations within each of the three main process caves - two Magnox fuel decanners and an advanced gas-cooled reactor fuel dismantler. The fuel route within the fuel handling plant is illustrated and described. ASPIC (Automated Sequence Package for Industrial Control) which was developed as a controller for the plant processes is described. (U.K.)

  1. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  2. Fundamental principles of failed fuel detection concepts on nuclear power units of WWER type

    International Nuclear Information System (INIS)

    Lusanova, L.; Miglo, V.; Slavyagin, P.

    2001-01-01

    The subject of the paper is the Russian failed fuel detection concept in both operating and shut down reactors. The philosophy for detection of fission products released from defective fuel during operation and sipping tests and using of these results for regulation of the radiological situation at the NPP during the next cycle is widely spread. In presented work such philosophy is applied to the shut down rectors. An option for sipping test performed in a mast of Refueling Machine (RM) using a wet-gas version of sipping test is briefly described. The use of the FFD method in RM mast allows combining the procedure of Fuel Assemblies (FA) tightness test with transport operation during reloading of the fuel from the core into the cooling pool. This reduces the time for reloading and transport operation with FA and increases the safety of reactor operation. The FFD method in RM mast has passed successful tests on Unit 4 at Balakovskaja NPP and it is expected to apply in other NPP unit with WWER-1000 reactors

  3. Flexibility of ADS for minor actinides transmutation in different two-stage PWR-ADS fuel cycle scenarios

    International Nuclear Information System (INIS)

    Zhou, Shengcheng; Wu, Hongchun; Zheng, Youqi

    2018-01-01

    Highlights: •ADS reloading scheme is optimized to raise discharge burnup and lower reactivity loss. •ADS is flexible to be combined with various pyro-chemical reprocessing technologies. •ADS is flexible to transmute MAs from different spent PWR fuels. -- Abstract: A two-stage Pressurized Water Reactor (PWR)-Accelerator Driven System (ADS) fuel cycle is proposed as an option to transmute minor actinides (MAs) recovered from the spent PWR fuels in the ADS system. At the second stage, the spent fuels discharged from ADS are reprocessed by the pyro-chemical process and the recovered actinides are mixed with the top-up MAs recovered from the spent PWR fuels to fabricate the new fuels used in ADS. In order to lower the amount of nuclear wastes sent to the geological repository, an optimized scattered reloading scheme for ADS is proposed to maximize the discharge burnup and lower the burnup reactivity loss. Then the flexibility of ADS for MA transmutation is evaluated in this research. Three aspects are discussed, including: different cooling time of spent ADS fuels before reprocessing, different reprocessing loss of spent ADS fuels, and different top-up MAs recovered from different kinds of spent PWR fuels. The ADS system is flexible to be combined with various pyro-chemical reprocessing technologies with specific spent fuels cooling time and unique reprocessing loss. The reduction magnitudes of the long-term decay heat and radiotoxicity of MAs by transmutation depend on the reprocessing loss. The ADS system is flexible to transmute MAs recovered from different kinds of spent PWR fuels, regardless of UOX or MOX fuels. The reduction magnitudes of the long-term decay heat and radiotoxicity of different MAs by transmutation stay on the same order.

  4. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  5. 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)

  6. 188Re-loaded lipid nanocapsules as a promising radiopharmaceutical carrier for internal radiotherapy of malignant gliomas

    International Nuclear Information System (INIS)

    Allard, E.; Hindre, F.; Passirani, C.; Lemaire, L.; Benoit, J.P.; Lepareur, N.; Noiret, N.; Menei, P.

    2008-01-01

    Lipid nanocapsules (LNC) entrapping lipophilic complexes of 188 Re( 188 Re(S 3 CPh) 2 (S 2 CPh) [ 188 Re-SSS]) were investigated as a novel radiopharmaceutical carrier for internal radiation therapy of malignant gliomas. The present study was designed to evaluate the efficacy of intra-cerebral administration of 188 Re-SSS LNC by means of convection-enhanced delivery (CED) on a 9L rat brain tumour model. Female Fischer rats with 9L glioma were treated with a single injection of 188 Re-SSS LNC by CED 6days after cell implantation. Rats were put into random groups according to the dose infused: 12, 10, 8 and 3Gy in comparison with blank LNC, perrhenate solution (4Gy) and non-treated animals. The radionuclide brain retention level was evaluated by measuring 188 Re elimination in faeces and urine over 72h after the CED injection. The therapeutic effect of 188 Re-SSS LNC was assessed based on animal survival. CED of 188 Re perrhenate solution resulted in rapid drug clearance with a brain T 1/2 of 7h. In contrast, when administered in LNC, 188 Re tissue retention was greatly prolonged, with only 10% of the injected dose being eliminated at 72h. Rat median survival was significantly improved for the group treated with 8Gy 188 Re-SSS LNC compared to the control group and blank LNC-treated animals. The increase in the median survival time was about 80% compared to the control group; 33% of the animals were long-term survivors. The dose of 8Gy proved to be a very effective dose, between toxic (10-12Gy) and ineffective (3-4Gy) doses. These findings show that CED of 188 Re-loaded LNC is a safe and potent anti-tumour system for treating malignant gliomas. Our data are the first to show the in vivo efficacy of 188 Re internal radiotherapy for the treatment of brain malignancy. (orig.)

  7. 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)

  8. Immersed multipurpose device for the inspection of PWR spent pencils in a loop reloaded in the Osiris pool

    International Nuclear Information System (INIS)

    Farny, Gerard.

    1982-09-01

    The equipment for the examination of fuel cans of irradiated fuel pencils by eddy current testing in the pool is described. Despite a high residual power there are no mechanical strains for the cladding and no interference with cooling. A high detection sensitivity is obtained. Friction is eliminated by water cushions and speed of operation is limited only by safety and data acquisition [fr

  9. Cost Savings of Nuclear Power with Total Fuel Reprocessing

    International Nuclear Information System (INIS)

    Solbrig, Charles W.; Benedict, Robert W.

    2006-01-01

    The cost of fast reactor (FR) generated electricity with pyro-processing is estimated in this article. It compares favorably with other forms of energy and is shown to be less than that produced by light water reactors (LWR's). FR's use all the energy in natural uranium whereas LWR's utilize only 0.7% of it. Because of high radioactivity, pyro-processing is not open to weapon material diversion. This technology is ready now. Nuclear power has the same advantage as coal power in that it is not dependent upon a scarce foreign fuel and has the significant additional advantage of not contributing to global warming or air pollution. A jump start on new nuclear plants could rapidly allow electric furnaces to replace home heating oil furnaces and utilize high capacity batteries for hybrid automobiles: both would reduce US reliance on oil. If these were fast reactors fueled by reprocessed fuel, the spent fuel storage problem could also be solved. Costs are derived from assumptions on the LWR's and FR's five cost components: 1) Capital costs: LWR plants cost $106/MWe. FR's cost 25% more. Forty year amortization is used. 2) The annual O and M costs for both plants are 9% of the Capital Costs. 3) LWR fuel costs about 0.0035 $/kWh. Producing FR fuel from spent fuel by pyro-processing must be done in highly shielded hot cells which is costly. However, the five foot thick concrete walls have the advantage of prohibiting diversion. LWR spent fuel must be used as feedstock for the FR initial core load and first two reloads so this FR fuel costs more than LWR fuel. FR fuel costs much less for subsequent core reloads ( 6 /MWe. The annual cost for a 40 year licensed plant would be 2.5 % of this or less if interest is taken into account. All plants will eventually have to replace those components which become radiation damaged. FR's should be designed to replace parts rather than decommission. The LWR costs are estimated to be 2.65 cents/kWh. FR costs are 2.99 cents/kWh for the first

  10. 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

  11. In-core fuel management for the course on operational physics of power reactors

    International Nuclear Information System (INIS)

    Levine, S.H.

    1982-01-01

    The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

  12. German Approach to Spent Fuel Management

    International Nuclear Information System (INIS)

    Jussofie, A.; Graf, R.; Filbert, W.

    2010-01-01

    The management of spent fuel was based on two powerful columns until 30 June 2005, i. e. reprocessing and direct disposal. After this date any delivery of spent fuel to reprocessing plants was prohibited so that the direct disposal of unreprocessed spent fuel is the only available option in Germany today. The main steps of the current concept are: (i) Intermediate storage of spent fuel, which is the only step in practice. After the first cooling period in spent fuel storage pools it continues into cask-receiving dry storage facilities. Identification of casks, 'freezing' of inventories in terms of continuity of knowledge, monitoring the access to spent fuel, verifying nuclear material movements in terms of cask transfers and ensurance against diversion of nuclear material belong to the fundamental safeguards goals which have been achieved in the intermediate storage facilities by containment and surveillance techniques in unattended mode. (ii) Conditioning of spent fuel assemblies by separating the fuel rods from structural elements. Since the pilot conditioning facility in Gorleben has not yet come into operation, the underlying safeguards approach which focuses on safeguarding the key measurement points - the spent fuel related way in and out of the facility - has not been applied yet. (iii) Disposal in deep geological formations, but no decision has been made so far neither regarding the location of a geological repository nor regarding the safeguards approach for the disposal concept of spent fuel. The situation was complicated by a moratorium which suspended the underground exploration of the Gorleben salt dome as potential geological repository for spent fuel. The moratorium expires in October 2010. Nevertheless, considerable progress has been made in the development of disposal concepts. According to the basic, so-called POLLUX (registered) -concept spent fuel assemblies are to be conditioned after dry storage and reloaded into the POLLUX (registered) -cask

  13. The exited hexagon reloaded

    Energy Technology Data Exchange (ETDEWEB)

    Bartels, Joachim [Hamburg Univ. (Germany). II. Inst. fuer Theoretische Physik; Kotanski, Jan; Schomerus, Volker; Sprenger, Martin [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany)

    2013-11-15

    This work revisits the computation of six-gluon scattering amplitudes in the high energy limit of strongly coupled N=4 supersymmetric Yang-Mills theory. It is based on previous studies in which we showed that the amplitude simplifies in the Regge regime and outlined an efficient computational scheme. By exploiting a symmetry of the underlying equations we are now able to argue that a term we had seen in preliminary numerical studies must vanish identically. The derived formula for the Regge limit of the 6-gluon scattering amplitude at strong coupling differs from the one we had conjectured previously.

  14. Reloading society [Editorial

    OpenAIRE

    Duca, Edward; Camilleri, Cassi

    2016-01-01

    Innovation breeds change, and with the current state of affairs change is in the pipeline. The tricky part is determining the right alterations for society— with research playing a huge role. We start with our cover story, introducing a new and innovative way to raise funds for charity while urging altruistic runners to perform better. Dr Franco Curmi and a diverse team (pg. 32) have designed a baton connected to social media that is carried by the jogger. A community of online supporters can...

  15. The exited hexagon reloaded

    International Nuclear Information System (INIS)

    Bartels, Joachim

    2013-11-01

    This work revisits the computation of six-gluon scattering amplitudes in the high energy limit of strongly coupled N=4 supersymmetric Yang-Mills theory. It is based on previous studies in which we showed that the amplitude simplifies in the Regge regime and outlined an efficient computational scheme. By exploiting a symmetry of the underlying equations we are now able to argue that a term we had seen in preliminary numerical studies must vanish identically. The derived formula for the Regge limit of the 6-gluon scattering amplitude at strong coupling differs from the one we had conjectured previously.

  16. Romanian nuclear fuel program: past, present and future

    International Nuclear Information System (INIS)

    Budan, O.; Rotaru, I.; Galeriu, C.A.

    1997-01-01

    and authorized CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle of January 1997 with fuel produced by the Romanian fuel plant. The quality evaluation of the 'pre-1990' fuel started in April 1996 and was performed by the Nuclear Fuel Plant (FCN) Pitesti, under the supervision of the Nuclear Power Group (GEN) - a distinct department of RENEL. The future prospect and trend of the Romanian Nuclear Fuel Program are also presented in this paper

  17. Fuel management optimization in pressure water reactors with hexagonal geometry using hill climbing method

    International Nuclear Information System (INIS)

    Andres Diaz, J.; Quintero, Ruben; Melian, Manuel; Rosete, Alejandro

    2000-01-01

    In this work the general-purpose optimization method, Hill Climbing, was applied to the Fuel Management Optimization problem in PWR reactors, WWER type. They were carried out a series of experiments in order to study the performance of Hill Climbing. It was proven two starting point for initialize the search: a reload configuration by project and a reload configuration generated with the application of a minimal knowledge of the problem. It was also studied the effect of imposing constraints based on the physics of the reactor in order to reduce the number of possible solutions to be generated. The operator used in Hill Climbing was defined as a binary exchange of fuel assemblies. For the simulation of each generated configuration, the tridimensional simulator program SPPS-1 was used. It was formulated an objective function with power peaking constraint to guide the search. As results, a methodology ws proposed for the In-core Fuel Management Optimization in hexagonal geometry, and the feasibility of the application of the Hill Climbing to this type of problem was demonstrated. (author)

  18. Efficiency improvement of nuclear power plant operation: the significant role of advanced nuclear fuel technologies

    International Nuclear Information System (INIS)

    Velde Van de, A.; Burtak, F.

    2001-01-01

    Due to the increased liberalisation of the power markets, nuclear power generation is being exposed to high cost reduction pressure. In this paper we highlight the role of advanced nuclear fuel technologies to reduce the fuel cycle costs and therefore increase the efficiency of nuclear power plant operation. The key factor is a more efficient utilisation of the fuel and present developments at Siemens are consequently directed at (i) further increase of batch average burnup, (ii) improvement of fuel reliability, (iii) enlargement of fuel operation margins and (iv) improvement of methods for fuel design and core analysis. As a result, the nuclear fuel cycle costs for a typical LWR have been reduced during the past decades by about US$ 35 million per year. The estimated impact of further burnup increases on the fuel cycle costs is expected to be an additional saving of US$10 - 15 million per year. Due to the fact that the fuel will operate closer to design limits, a careful approach is required when introducing advanced fuel features in reload quantities. Trust and co-operation between the fuel vendors and the utilities is a prerequisite for the common success. (authors)

  19. Fuel cycle and waste management. 2. Design of a BWR Core with Over-moderated MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Francois, J.L.; Del Campo, C. Martin

    2001-01-01

    The use of uranium-plutonium mixed-oxide (MOX) fuel in light water reactors is a current practice in several countries. Generally one-third of the reactor core is loaded with MOX fuel assemblies, and the other two-thirds is loaded with uranium assemblies. Nevertheless, the plutonium utilization could be more effective if the full core could be loaded with MOX fuel. In this work, the design of a boiling water reactor (BWR) core fully loaded with over-moderated MOX fuel designs was investigated. In previous work, the design of over-moderated BWR MOX fuel assemblies based on a 10 x 10 lattice was presented; these designs improve the neutron spectrum and the plutonium consumption rate, compared with standard MOX assemblies. To increase the moderator-to-fuel ratio (MFR), two approaches were followed. In the first approach, 8 or 12 fuel rods were replaced by water rods in the 10x10 assembly, which increased the MFR from 1.9 to 2.2 and 2.4, respectively. These designs are called MOX-8WR and MOX-12WR, respectively, in this paper. In the second approach, an 11 x 11 lattice with 24 water rods (11 x 11-24WR) was designed, which is a design with a number of active fuel rods (88) very close to the standard MOX assembly (91). The fuel rod diameter is smaller to preserve the assembly dimensions, and in this last case, the MFR is 2.4. The calculations were performed with the CM-PRESTO three-dimensional steady-state simulator. The nuclear data banks were generated with the HELIOS system, and they were processed by TABGEN to produce tables of nuclear cross sections depending on burnup, void, and exposure weighted void (void history), which are used by CM-PRESTO. One base reload pattern was designed for a BWR/5 rated at 1931 MW(thermal), to be used with the different over-moderated assembly designs. The reload pattern has 112 fresh fuel assemblies (FFAs) out of a total of 444 fuel assemblies and was simulated during 20 cycles with the Haling strategy, until an equilibrium cycle of

  20. GENUSA Fuel Evolution

    Energy Technology Data Exchange (ETDEWEB)

    Choithramani, Sylvia; Malpica, Maria [ENUSA Industrias Avanzadas, GENUSA, Josefa Valcarcel, 26 28027 Madrid (Spain); Fawcett, Russel [Global Nuclear Fuel (United States)

    2009-06-15

    surface specifications to add PCI margin; - Introduction of a debris filter, applied as a standard feature to 10x10 GE14, and as an optional feature in 9x9 fuel, to address debris fretting, as well as advancements to debris filters to achieve even better resistance to debris ingress. GENUSA has always taken the necessary steps to assure the infrastructure and technology are in place to support each product or potential product introduction program. This paper will describe these steps and the evolution of the GENUSA delivered product in Europe starting with the first Garona reload product and finish with a slight description of how our latest product, GNF2, was born. This will include how GENUSA opened to the European market and all the different products that GENUSA has offered and offers nowadays. (authors)

  1. Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

    International Nuclear Information System (INIS)

    Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

    2000-01-01

    The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

  2. 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

  3. Data mining in the study of nuclear fuel cells

    International Nuclear Information System (INIS)

    Medina P, J. A.; Ortiz S, J. J.; Castillo, A.; Montes T, J. L.; Perusquia, R.

    2015-09-01

    In this paper is presented a study of data mining application in the analysis of fuel cells and their performance within a nuclear boiling water reactor. A decision tree was used to fulfill questions of the type If (condition) and Then (conclusion) to classify if the fuel cells will have good performance. The performance is measured by compliance or not of the cold shutdown margin, the rate of linear heat generation and the average heat generation in a plane of the reactor. It is assumed that the fuel cells are simulated in the reactor under a fuel reload and rod control patterns pre designed. 18125 fuel cells were simulated according to a steady-state calculation. The decision tree works on a target variable which is one of the three mentioned before. To analyze this objective, the decision tree works with a set of attribute variables. In this case, the attributes are characteristics of the cell as number of gadolinium rods, rods number with certain uranium enrichment mixed with a concentration of gadolinium, etc. The found model was able to predict the execution or not of the shutdown margin with a precision of around 95%. However, the other two variables showed lower percentages due to few learning cases of the model in which these variables were or were not achieved. Even with this inconvenience, the model is quite reliable and can be used in way coupled in optimization systems of fuel cells. (Author)

  4. 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

  5. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  6. Nencki Affective Word List (NAWL): the cultural adaptation of the Berlin Affective Word List-Reloaded (BAWL-R) for Polish.

    Science.gov (United States)

    Riegel, Monika; Wierzba, Małgorzata; Wypych, Marek; Żurawski, Łukasz; Jednoróg, Katarzyna; Grabowska, Anna; Marchewka, Artur

    2015-12-01

    In the present article, we introduce the Nencki Affective Word List (NAWL), created in order to provide researchers with a database of 2,902 Polish words, including nouns, verbs, and adjectives, with ratings of emotional valence, arousal, and imageability. Measures of several objective psycholinguistic features of the words (frequency, grammatical class, and number of letters) are also controlled. The database is a Polish adaptation of the Berlin Affective Word List-Reloaded (BAWL-R; Võ et al., Behavior Research Methods 41:534-538, 2009), commonly used to investigate the affective properties of German words. Affective normative ratings were collected from 266 Polish participants (136 women and 130 men). The emotional ratings and psycholinguistic indexes provided by NAWL can be used by researchers to better control the verbal materials they apply and to adjust them to specific experimental questions or issues of interest. The NAWL is freely accessible to the scientific community for noncommercial use as supplementary material to this article.

  7. Introduction of fuel GE14 in the nuclear power plant of Laguna Verde for the extended increase of power; Introduccion del combustible GE14 en la central nuclear Laguna Verde para el aumento de potencia extendido

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Vargas A, A. F.; Cardenas J, J. B.; Contreras C, P. [CFE, Central Nuclear Laguna Verde, Subgerencia de Ingenieria, Carretera Veracruz-Medellin Km. 7.5 (Mexico)]. e-mail: natividad.hernandez@cfe.gob.mx

    2008-07-01

    The project of extended increase of power responds to a necessity of electrical energy in the country, increasing the thermal exit of the reactors of the nuclear power plant of Laguna Verde of 2027 MWt to 2317 MWt. In order to support this transition, changes will make in the configuration of the reactor core and in the operation strategies of the cycle, also they will take initiatives to optimize the economy in fuel cycle. At present in both reactors of the nuclear plant of Laguna Verde fuel GE12 is used. The fuel GE14 presents displays with respect to the GE12, some improvements in the mechanical design and consequently in its performance generally. Between these improvements we can mention: 1. Spacers of high performance. 2. Shielding with barrier. 3. Filter for sweepings {sup d}ebris{sup a}nd 4. Fuel rods of minor partial length. The management of nuclear power plants has decided to introduce the use of fuel GE14 in Laguna Verde in the reload 14 for Unit 1 and of the reload 10 for Unit 2. The process of new introduction fuel GE14 consists of two stages, first consists on subjecting the one new design of fuel to the regulator organism in the USA: Nuclear Regulatory Commission, in Mexico the design must be analyzed and authorized by the National Commission of Nuclear Safety and Safeguards, for its approval of generic form, by means of the demonstration of the fulfillment with the amendment 22 of GESTAR II, the second stage includes the specific analyses of plant to justify the use of the new fuel design in a reload core. The nuclear plant of Laguna Verde would use some of the results of the security analyses that have been realized for the project of extended increase of power with fuel GE14, to document the specific analyses of plant with the new fuel design. The result of the analyses indicates that the reload lots are increased of 116-120 assemblies in present conditions (2027 MWt) to 140-148 assemblies in conditions of extended increase of power (2317 MWt

  8. Establishing the long-term fuel management scheme using point reactivity model

    International Nuclear Information System (INIS)

    Park, Yong-Soo; Kim, Jae-Hak; Lee, Young-Ouk; Song, Jae-Woong; Zee, Sung-Kyun

    1994-01-01

    A new approach to establish the long-term fuel management scheme is presented in this paper. The point reactivity model is used to predict the core average reactivity. An attempt to calculate batchwise power fraction is introduced through the two-dimensional nodal power algorithm based on the modified one-group diffusion equation and the number of fuel assemblies on the core periphery. Suggested is an empirical formula to estimate the radial leakage reactivity with ripe core design experience reflected. This approach predicts the cycle lengths and the discharge burnups of individual fuel batches up to an equilibrium core when the proper input data such as batch enrichment, batch size, type and content of burnable poison and reloading strategies are given. Eight benchmark calculations demonstrate that the new approach used in this study is reasonably accurate and highly efficient for the purpose of scoping calculation when compared with design code predictions. (author)

  9. Modeling gadolinium-bearing fuel in Ringhals PWRs using CASMO/SIMULATE

    International Nuclear Information System (INIS)

    Kurcyusz, E.

    1993-01-01

    Ringhals units 2, 3, and 4 are Westinghouse three-loop, 157-assembly pressurized water reactors (PWRs) operated by Vattenfall. Originally, all three reactors were loaded in an out-in scheme using reload fuel without burnable poisons. In recent cycles, gadolinium-bearing fuel was introduced to enable a low-leakage loading pattern and minimize fuel cycle costs. This paper focuses on the Fragema 17 x 17 AFA design with peripheral gadolinium rods loaded in units 3 and 4. The Ringhals units are modeled using the Studsvik core management system, consisting of the CASMO-3 transport theory lattice physics code,and the SIMULATE-3 advanced nodal reactor analysis code. The results of the studies verifying the accuracy of CASMO-3/SIMULATE-3 on the assemblies with peripheral gadolinium rods are presented in this paper. The verification was carried out against CASMO-3 color-set calculations and measured reactor data

  10. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  11. Technical assistance in relationship with the reloading analysis of the Laguna Verde Unit 2 reactor. Executive abstract

    International Nuclear Information System (INIS)

    Alonso V, G.; Castro B, M.; Gallegos E, R.; Hernandez L, H.; Montes T, J.L.; Ortiz S, J. J.; Perusquia C, R.

    1993-11-01

    The objective of the report was to carry out a comparative analysis of costs of energy generation among the designs GE9B of General Electric, 9X9-IX of SIEMENS and SVEA-96 of ABB ATOM, proposed to be used as recharge fuel in the Unit 2 of the Laguna Verde Nuclear Power station. (Author)

  12. Sequence assembly

    DEFF Research Database (Denmark)

    Scheibye-Alsing, Karsten; Hoffmann, S.; Frankel, Annett Maria

    2009-01-01

    Despite the rapidly increasing number of sequenced and re-sequenced genomes, many issues regarding the computational assembly of large-scale sequencing data have remain unresolved. Computational assembly is crucial in large genome projects as well for the evolving high-throughput technologies and...... in genomic DNA, highly expressed genes and alternative transcripts in EST sequences. We summarize existing comparisons of different assemblers and provide a detailed descriptions and directions for download of assembly programs at: http://genome.ku.dk/resources/assembly/methods.html....

  13. Genome Sequencing

    DEFF Research Database (Denmark)

    Sato, Shusei; Andersen, Stig Uggerhøj

    2014-01-01

    The current Lotus japonicus reference genome sequence is based on a hybrid assembly of Sanger TAC/BAC, Sanger shotgun and Illumina shotgun sequencing data generated from the Miyakojima-MG20 accession. It covers nearly all expressed L. japonicus genes and has been annotated mainly based on transcr......The current Lotus japonicus reference genome sequence is based on a hybrid assembly of Sanger TAC/BAC, Sanger shotgun and Illumina shotgun sequencing data generated from the Miyakojima-MG20 accession. It covers nearly all expressed L. japonicus genes and has been annotated mainly based...

  14. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  15. Experience of RIA safety analyses performance for NPP Temelin core arranged with TVSA-T fuel assemblies

    International Nuclear Information System (INIS)

    Kryukov, S.A.; Lizorkin, M.P.

    2010-01-01

    The contents of the presentation are as follows: 1. Definition of categories for initiating events; 2. Acceptance criteria for safety assessment; 3. Main aspects of safety assessment methodology; 4. Main stages of calculation analysis; 5. Interface with other parts of the core design; 6. Codes used for calculation; 6.1 Main performances of code package TIGR-1; 6.2 Main performances of code BIPR-7A; 7. TIGR-1 accounting of design margins in calculation of fuel rod powers; 8. Peculiar features of Instrumentation and Control System for Temelin NPP; 9. Calculations; 10. Checklist of margin data important for reload safety assessment. (P.A.)

  16. User's guide for the REBUS-3 fuel cycle analysis capability

    International Nuclear Information System (INIS)

    Toppel, B.J.

    1983-03-01

    REBUS-3 is a system of programs designed for the fuel-cycle analysis of fast reactors. This new capability is an extension and refinement of the REBUS-3 code system and complies with the standard code practices and interface dataset specifications of the Committee on Computer Code Coordination (CCCC). The new code is hence divorced from the earlier ARC System. In addition, the coding has been designed to enhance code exportability. Major new capabilities not available in the REBUS-2 code system include a search on burn cycle time to achieve a specified value for the multiplication constant at the end of the burn step; a general non-repetitive fuel-management capability including temporary out-of-core fuel storage, loading of fresh fuel, and subsequent retrieval and reloading of fuel; significantly expanded user input checking; expanded output edits; provision of prestored burnup chains to simplify user input; option of fixed-or free-field BCD input formats; and, choice of finite difference, nodal or spatial flux-synthesis neutronics in one-, two-, or three-dimensions

  17. An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

    2012-07-01

    The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

  18. IFPE/IFA-533, Fuel Thermal Behaviour at High Burnup, Halden Reactor

    International Nuclear Information System (INIS)

    Gyori, Cs.; Turnbull, J.A.

    1997-01-01

    Description: After twelve years irradiation in the Halden Boiling Water Reactor two fuel rods (Rod 807 and Rod 808) were re-instrumented with fuel centre thermocouples and reloaded into the reactor in order to investigate the fuel thermal behaviour at high burnup. The fuel rods were pre-irradiated with four other rods in the upper cluster of IFA-409 (IFA=Instrumented Fuel Assembly) from May 1973 to June 1985. After base irradiation the four neighbouring rods were re-instrumented with pressure transducers and ramp tested in IFA-535.5 and IFA-535.6 providing useful data about fission gas release (FGR) presented in the Fuel Performance Database as well (Ref. 1). The two rods re-instrumented with fuel centre thermocouples have been irradiated as IFA-533.2 from April 1992. As the irradiation history of IFA-533.2 in the first months was very similar to the history of the ramp tests, the fuel temperature and FGR data measured in the different IFAs can complement each other, although the fuel-cladding gap sizes were slightly different and due to re-instrumentation the internal gas conditions were also dissimilar

  19. GNPS 18-months fuel cycles core thermal hydraulic design

    International Nuclear Information System (INIS)

    Liu Changwen; Zhou Zhou

    2002-01-01

    GNPS begins to implement the 18-month fuel cycles from the initial annual reload at cycle 9, thus the initial core thermal hydraulic design is not valid any more. The new critical heat flux (CHF) correlation, FC, which is developed by Framatome, is used in the design, and the generalized statistical methodology (GSM) instead of the initial deterministic methodology is used to determine the DNBR design limit. As the AFA 2G and AFA 3G are mixed loaded in the transition cycle, it will result that the minimum DNBR in the mixed core is less than that of AFA 3G homogenous core, the envelop mixed core DNBR penalty is given. Consequently the core physical limit for mixed core and equilibrium cycles, and the new over temperature ΔT overpower ΔT are determined

  20. Development of packagings for 'MONJU' blanket fuel assemblies

    International Nuclear Information System (INIS)

    Shibata, Kan; Ouchi, Yuichiro; Matsuzaki, Masaaki; Okuda, Yoshihisa

    1995-01-01

    Blanket assemblies for prototype Fast Breeder Reactor 'MONJU' are made at commercial fuel fabrication plants capable of handling deplete Uranium in Japan. For the purpose of transport the assemblies are inserted into a packaging that is set horizontally at the fabrication plants because of compatibility with equipment installed at the plants. On the other hand, the assemblies must be taken out from the packaging set vertically at 'MONJU' due to compatibility. For this reason development of a new packaging, which makes it possible to take assemblies in and out both horizontally and vertically, is needed to carry out transport of assemblies for reload. The development and fabrication of the packagings, taking about two years, were completed in March 1995. The packagings were used in transport of assemblies in June 1995 for the first change. This report introduces the outline of the packaging and confirmation tests done in the process of development. (author)

  1. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1980-09-01

    Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U.

  2. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Third semiannual report, January-June 1980

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1980-09-01

    Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the work scope of this program one of these concepts is to be selected for demonstration in a commercial power reactor. It was decided to demonstrate Zr-liner in 132 bundles which have liners of either crystal-bar zirconium or of low-oxygen sponge zirconium in the reload for Quad Cities Unit 2, Cycle 6. Irradiation testing or barrier fuel was continued, and the superior PCI resistance of Zr-liner fuel was further substantiated in the current report period. Furthermore, an irradiation experiment in which Zr-liner fuel, having a deliberately fabricated cladding perforation, was operated at a linear heat generation rate of 35 kW/m to a burnup of approx. 3 MWd/kg U showed no unusual signs of degradation compared with a similarly defected reference fuel rod. Four lead test assemblies of barrier fuel (two of Zr-liner and two of Cu-barrier), presently under irradiation in Quad Cities Unit 1, have achieved a burnup of 11 MWd/kg U

  3. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  6. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  7. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  8. Caramel fuel for research reactors: experience acquired in the fabrication, monitoring and irradiation of Osiris core

    International Nuclear Information System (INIS)

    Contenson, Ghislain de; Foulquier, Henri; Trotabas, Maria; Vignesoult, Nicole; Cerles, J.-M.; Delafosse, Jacques.

    1981-06-01

    A plate type nuclear fuel (Caramel fuel) has been developed in France in the framework of the various activities pursued in the design, fabrication and development of nuclear fuels by the CEA. This fuel can be adapted to various different categories of water cooled reactor (power reactors, marine propulsion reactors, urbain heating reactors, research reactors). The successful work conducted in this field led the realization of a complete core and reloads for the high performance research reactor, Osiris, at Saclay. The existing highly enriched U-Al alloy fuel was replaced by a non-proliferating low enrichment (7%) caramel fuel. This new core has been operating successfully since january 1980. A brief description of Caramel and its main advantages is given. The way in which it is fabricated is described together with the quality controls to which it is subjected. The qualification program and the main results deduced from it are also presented. The program used to monitor its in-pile behavior is described. The essential purpose of this program is to ensure the high performance of the fuel under irradiation. The successful operation of Osiris, which terminated 11 irradiation cycles on the 21st of April 1981 confirmed the correctness of the decisions made and the excellent performance that could be achieved with these fuel elements under the severe conditions encountered in a high performance research reactor [fr

  9. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  10. Study in stationary state of the subcriticality of intermediate configurations of core in the reloading process of a BWR

    International Nuclear Information System (INIS)

    Hernandez, J.L.; Montes, J.L.; Perusquia, R.; Ortiz, J.J.

    2006-01-01

    In this work is carried out the simulation in three dimensions with the COREMASTER-PRESTO code, of the behavior of the reactor core in different stages of the change process of fuel assemblies. To carry out the simulation, this code requires of a database of nuclear parameters that includes those that can associate to the areas of an assemblies that they don't contain fuel and in its place there is moderator. These nuclear parameters are calculated with the AURORA-HELIOS-ZENITH-TABGEN system. One of the approaches that were carried out consisted on designing a 'water assemble', that is to say, an axial arrangement of 25 'water cells'. To obtain the appropriate 'water cell' its were carried out some selective test cases, since it presents in two cases the necessity to find an enough minimum value of fissile material for the correct execution of HELIOS, firstly, and later on COREMASTER-PRESTO. In the first case, the situation is solved when placing symmetrically 6 bars with natural uranium in the lateral areas of the cell; with that which the value of k inf of 0.1592 is obtained in the calculations with the HELIOS code in the cold condition to zero power (CZP), and 0% of vacuums. For the second case the cell includes symmetrically 28 bars with natural uranium, and the k inf value is 0.45290. These values are the maximum through the life of the 'cell.' As part of the activities that are developed during the fuel substitution, this the one of evaluating the subcriticality of the core each determined number of substitution movements. The obtained results when evaluating the k-effective in cold condition, in 5 different intermediate core configurations, as the loading process of the fuel advances are presented. To make the evaluation with CM-PRESTO in each configuration it was proceeded to complete the rest of the 444 assemblies with the one denominated 'water assemble'. In all the evaluated cases the subcriticality of the core was demonstrated in cold condition and with

  11. BWR simulation in a stationary state for the evaluation of fuel cell design; Simulacion de un reactor BWR en estado estacionario para la evaluacion del diseno de celdas de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J. L.; Ortiz S, J. J.; Perusquia del C, R.; Castillo M, A., E-mail: joseluis.montes@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this paper the simulation of a BWR in order to evaluate the performance of a set of fuel assemblies under stationary state in three dimensions (3-D) is presented. 15 cases selected from a database containing a total of 18225 cases are evaluated. The main selection criteria were based on the results of the design phase of the power cells in two dimensions (2-D) and 3-D initial study. In 2-D studies the parameters that were used to qualify and select the designs were basically the local power peaking factor and neutron multiplication factor of each fuel cell. In the initial 3-D study variables that defined the quality of results, and from which the selection was realized, are the margins to thermal limits of reactor operation and the value of the effective multiplication factor at the end of cycle operation. From the 2-D and 3-D results of the studies described a second 3-D study was realized, where the optimizations of the fuel reload pattern was carried out. The results presented in this paper correspond to this second 3-D study. It was found that the designs of the fuel cell they had a similar behavior to those provided by the fuel supplier of reference BWR. Particularly it noted the impact of reload pattern on the cold shut down margin. An estimate of the operation costs of reference cycle analyzed with each one designed reload batch was also performed. As a result a positive difference (gain) up to 10,347 M/US D was found. (Author)

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  13. The fabrication of nuclear fuel elements in Mexico

    International Nuclear Information System (INIS)

    Guerrero Morillo, H.L.

    1977-01-01

    The situation of nuclear electricity generation in Mexico in 1976 is described: two nuclear reactors were under construction but no definite programme on the type and start-up dates for the next power plants existed. However, the existence of a general plan on nuclear power plants is mentioned, which, according to the latest estimates, will provide 10,000MW installed by 1990. The national intention, as laid down in an appropriate Law, is to supply domestic nuclear fuel to the power reactors operating in the country, starting with the first reloading of the two BWRs at the first national station in Laguna Verde, required at the end of 1981 and 1982, respectively. Before this can be achieved and to provide the relatively small amounts of fuel elements for the two reactors, Mexico must adopt a strategy of fuel elements fabrication. The two main options are analysed: (1) to delay local fabrication until a national nuclear programme has been defined, meanwhile purchasing abroad the necessary initial cores and refuelling; (2) to start local fabrication of fuel elements as soon as possible in order to provide the first refuelling of the first unit of Laguna Verde, confronting the economic risks of such a decision with the advantages of immediate action. Both options are analysed in detail, comparing them especially from the economic point of view. Current information from potential licensors for design and manufacture are used in the analysis. (author)

  14. Statistical analysis in the design of nuclear fuel cells and training of a neural network to predict safety parameters for reactors BWR

    International Nuclear Information System (INIS)

    Jauregui Ch, V.

    2013-01-01

    In this work the obtained results for a statistical analysis are shown, with the purpose of studying the performance of the fuel lattice, taking into account the frequency of the pins that were used. For this objective, different statistical distributions were used; one approximately to normal, another type X 2 but in an inverse form and a random distribution. Also, the prediction of some parameters of the nuclear reactor in a fuel reload was made through a neuronal network, which was trained. The statistical analysis was made using the parameters of the fuel lattice, which was generated through three heuristic techniques: Ant Colony Optimization System, Neuronal Networks and a hybrid among Scatter Search and Path Re linking. The behavior of the local power peak factor was revised in the fuel lattice with the use of different frequencies of enrichment uranium pines, using the three techniques mentioned before, in the same way the infinite multiplication factor of neutrons was analyzed (k..), to determine within what range this factor in the reactor is. Taking into account all the information, which was obtained through the statistical analysis, a neuronal network was trained; that will help to predict the behavior of some parameters of the nuclear reactor, considering a fixed fuel reload with their respective control rods pattern. In the same way, the quality of the training was evaluated using different fuel lattices. The neuronal network learned to predict the next parameters: Shutdown Margin (SDM), the pin burn peaks for two different fuel batches, Thermal Limits and the Effective Neutron Multiplication Factor (k eff ). The results show that the fuel lattices in which the frequency, which the inverted form of the X 2 distribution, was used revealed the best values of local power peak factor. Additionally it is shown that the performance of a fuel lattice could be enhanced controlling the frequency of the uranium enrichment rods and the variety of the gadolinium

  15. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  16. Measurement of gamma ray from fuel of high temperature engineering test reactor. Method of measurement and results

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment] [and others

    2001-02-01

    To obtain information in the HTTR core directly, gamma ray from fuel blocks was measured when fuel blocks were discharged from the core and reloaded to the core. Gamma ray was measured using GM detector, CZT semiconductor detector installed in a door valve and area monitors installed in a stand pipe compartment. The measurement was carried out for 20 fuel blocks in 4 columns considering the symmetry of uranium enrichment distribution in the core. Relative axial distribution in each column obtained by the GM detector and CZT detector agreed with calculated results. However, calculation values showed higher values than measured values in upper region of the core, lower those in lower region of the core. The axial distributions were also evaluated by the area monitors. The measured values agreed with calculated values. It became clear that it was possible to obtain the data inside the core by this method. (author)

  17. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    International Nuclear Information System (INIS)

    Hampel, Gabriele; Cordes, Harro; Ebbinghaus, Kurt; Haferkamp, Dirk

    1998-01-01

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  18. The potential for expert system support in solving the pressurized water reactor fuel shuffling problem

    International Nuclear Information System (INIS)

    Rothleder, B.M.; Poetschat, G.R.; Faught, W.S.; Eich, V.J.

    1988-01-01

    The fuel shuffling problem is posed by the need to reposition partially burned assemblies to achieve minimum X-Y pin power peaks reload cycles of pressurized water reactors. This problem is a classic artificial intelligence (AI) problem and is highly suitable for AI expert system solution assistance, in contrast to the conventional solution, which ultimately depends solely on trial and error. Such a fuel shuffling assistant would significantly reduce engineering and computer execution time for conventional loading patterns and, much more importantly, even more significantly for low-leakage loading patterns. A successful hardware/software demonstrator has been introduced, paving the way for development of a broadly expert system program. Such a program, upon incorporating the recently developed technique perverse depletion, would provide a directed path for solving the low-leakage problem

  19. Engineering program in order to increase the irradiated fuel storage capacity in pool facilities of Juragua

    International Nuclear Information System (INIS)

    Rodriguez R, J.

    1996-01-01

    In 1993, a technical program in the spent fuel storage area of Nuclear Plant Juragua was launched. Such a program tries to carry out an engineering assessment of the possibility of increasing the spent fuel storage capacity in pool storage facilities by using high density racks (re-racking) instead of the original (non-compact) ones. The purpose of the above-mentioned program is to evaluate possible solutions that can be applied to the construction works prior to plant operation. The first stage of the program for the 1994-95 period is an ongoing Engineering-Economic Feasibility Study (EEFS), which endeavors to examine the capabilities of the reloading pool in Unit-1 Reactor building and long-term storage pool in auxiliary building in high density storage conditions. Technical details of the EEFS and reached results and difficulties are described. (author). 5 refs., 2 figs

  20. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-01-01

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  2. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  3. Nuclear fuel rod loading apparatus

    International Nuclear Information System (INIS)

    King, H.B.; Macivergan, R.; Mckenzie, G.W.

    1980-01-01

    An apparatus incorporating a microprocessor control is provided for automatically loading nuclear fuel pellets into fuel rods commonly used in nuclear reactor cores. The apparatus comprises a split ''v'' trough for assembling segments of fuel pellets in rows and a shuttle to receive the fuel pellets from the split ''v'' trough when the two sides of the split ''v'' trough are opened. The pellets are weighed while in the shuttle, and the shuttle then moves the pellets into alignment with a fuel rod. A guide bushing is provided to assist the transfer of the pellets into the fuel rod. A rod carousel which holds a plurality of fuel rods presents the proper rod to the guide bushing at the appropriate stage in the loading sequence. The bushing advances to engage the fuel rod, and the shuttle advances to engage the guide bushing. The pellets are then loaded into the fuel rod by a motor operated push rod. The guide bushing includes a photocell utilized in conjunction with the push rod to measure the length of the row of fuel pellets inserted in the fuel rod

  4. Fuel operation of EDF nuclear fleet presentation of the centralized organization for operational engineering at the nuclear generation division

    International Nuclear Information System (INIS)

    Paulin, Ph.

    2006-01-01

    The main feature of EDF Nuclear Fleet is the standardization, with 'series' of homogeneous plants (same equipment, fuel and operation technical documents). For fuel operation, this standardization is related to the concept of 'fuel management scheme' (typical fuel reloads with fixed number and enrichment of fresh assemblies) for a whole series of plants. The context of the Nuclear Fleet lead to the choice of a centralized organization for fuel engineering at the Nuclear Generation Division (DPN), located at UNIPE (National Department for Fleet Operation Engineering) in Lyon. The main features of this organization are the following: - Centralization of the engineering activities for fuel operation support in the Fuel Branch of UNIPE, - Strong real-time link with the nuclear sites, - Relations with various EDF Departments in charge of design, nuclear fuel supply and electricity production optimization. The purposes of the organization are: - Standardization of operational engineering services and products, - Autonomy with independent methods and computing tools, - Reactivity with a technical assistance for sites (24 hours 'hot line'), - Identification of different levels (on site and off site) to solve core operation problems, - Collection, analysis and valorization of operation feedback, - Contribution to fuel competence global management inside EDF. This paper briefly describes the organization. The main figures of annual engineering production are provided. A selection of examples illustrates the contribution to the Nuclear Fleet performance. (authors)

  5. Nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    1982-01-01

    This film for a general audience deals with nuclear fuel waste management in Canada, where research is concentrating on land based geologic disposal of wastes rather than on reprocessing of fuel. The waste management programme is based on cooperation of the AECL, various universities and Ontario Hydro. Findings of research institutes in other countries are taken into account as well. The long-term effects of buried radioactive wastes on humans (ground water, food chain etc.) are carefully studied with the help of computer models. Animated sequences illustrate the behaviour of radionuclides and explain the idea of a multiple barrier system to minimize the danger of radiation hazards

  6. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

    2011-01-01

    -standard position of fuel assembly components during unloading from the core. In addition all 42 assemblies were subject to the standard leak testing process with all found to be hermetically sealed. Six fuel assemblies of this Westinghouse reload batch were then subjected to a more extensive inspection program similar to what was done during the LTA program. Detailed results and concluding remarks from the post irradiation examination is provided in this paper. Westinghouse has now completed manufacturing and delivery of three regions of the Westinghouse WWER-1000 design to the South Ukraine NPP. Manufacturing of these fuel campaigns has gone very well without major issues, and the production of the WWER-1000 design has been integrated successfully with the other product lines in Sweden. In the second half of 2012 the first region of fuel to the Zaporizhzhya NPP will be delivered. (authors)

  7. Nuclear Fuel Design Considerations for the 1990s

    International Nuclear Information System (INIS)

    Stucker, David L.

    1993-01-01

    Nuclear fuel for many of today's operating Ness's was designed based on the expectation of annual fuel cycles, plutonium recycle, low cost uranium commodities, and discharge burnups of about 33 GW D/Mtu. The original PWR Ness designers envisioned equilibrium annual cycles with negative moderator feedback at all times. The annual cycle and low discharge burnup could be easily achieved without the use of burnable absorbers in all but the first fuel cycle using classical out-in core loading techniques. Fuel assembly insert burnable absorbers were developed to maintain negative moderator feedback for first cycles but were not optimized for use in reload cycles due to their perceived limited application. The plutonium recycle assumption has proven to be one with major design implications. Low discharge burnups to maximize the fissile content of the total plutonium generated, relatively low H/U ratios to promote plutonium breeding, spent fuel storage capacity sized by cooling requirements not plant lifetime, and less importance placed upon the use of parasitic materials within the reactor volume are all outcomes of the plutonium recycle design assumption. Historically, the plutonium recycle assumption has proven to be an unfortunate one in that fuel arrays and Ness hardware were designed and compromised to accommodate a fuel cycle alternative that has seen little economic or political success. Utility customers in the 1990s require ever-increasing fuel discharge burnup and hot residence time, continuing thermal margin improvement, efficient burnable absorbers, continued reductions in fuel cycle, operation and maintenance costs, and reductions in worker radiation exposure. In addition, because the costs associated with fuel rod defects are extremely high, both in currency and worker exposure, all of these competitive pressures come with the foremost requirement of defect-free operation. Fuel assembly vendors have responded to these competitive pressures with advanced

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  9. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  10. Elements of nuclear reactor fueling theory

    International Nuclear Information System (INIS)

    Egan, M.R.

    1984-01-01

    Starting with a review of the simple batch size effect, a more general theory of nuclear fueling is derived to describe the behavior and physical requirements of operating cycle sequences and fueling strategies having practical use in the management of nuclear fuel. The generalized theory, based on linear reactivity modeling, is analytical and represents the effects of multiple-stream, multiple-depletion-batch fueling configurations in systems employing arbitrary, non-integer batch size strategies, and containing fuel with variable energy generation rates. Reactor operating cycles and cycle sequences are represented with realistic structure that includes the effects of variable cycle energy production, cycle lengths, end-of-cycle operating extensions and maneuvering allowances. Results of the analytical theory are first applied to the special case of degenerate equilibrium cycle sequences, yielding several fundamental principles related to the selection of refueling strategy, and which govern fueling decisions normally made by the fuel manager. It is also demonstrated in this application that the simple batch size effect is not valid for non-integer fueling strategies, even in the simplest sequence configurations, and that it systematically underestimates the fueling requirements of degenerate sequences in general

  11. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Anniel, Marc; Bordy, Michel-Aristide

    2009-01-01

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  12. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  13. Fuel management

    International Nuclear Information System (INIS)

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  14. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  15. Fuel cycles

    International Nuclear Information System (INIS)

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  16. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  17. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  18. Spent Nuclear Fuel (SNF) Removal Campaign Plan

    International Nuclear Information System (INIS)

    PAJUNEN, A.L.

    2000-01-01

    The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage

  19. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  20. Fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

  1. Non-standard constraints within In-Core Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    Maldonado, G.I. [University of Cincinnati, P.O. Box 210072, Cincinnati, OH 45221-0072 (United States); Torres, C. [Comision Federal de Electricidad, Gestion de Combustible, Mexico, D.F. (Mexico); Marrote, G.N.; Ruiz U, V. [Global Nuclear Fuel, Americas, LLC, PO Box 780, M/C A16, Wilmington, NC28402 (United States)]. e-mail: Ivan.Maldonado@uc.edu

    2004-07-01

    Recent advancements in the area of nuclear fuel management optimization have been considerable and widespread. Therefore, it is not surprising that the design of today's nuclear fuel reloads can be a highly automated process that is often accompanied by sophisticated optimization software and graphical user interfaces to assist core designers. Most typically, among other objectives, optimization software seeks to maximize the energy efficiency of a fuel cycle while satisfying a variety of safety, operational, and regulatory constraints. Concurrently, the general industry trend continues to be one of pursuing higher generating capacity (i.e., power up-rates) alongside cycle length extensions. As these increasingly invaluable software tools and ambitious performance goals are pursued in unison, more aggressive core designs ultimately emerge that effectively minimize the margins to limits and, in some cases, may turn out less forgiving or accommodating to changes in underlying key assumptions. The purpose of this article is to highlight a few 'non-standard', though common constraints that can affect a BWR core design but which are often difficult, if not impossible, to implement into an automated setting. In a way, this article indirectly emphasizes the unique and irreplaceable role of the experienced designer in light of 'real life' situations. (Author)

  2. Experimental conditions at Osiris with the new CARAMEL fuel

    International Nuclear Information System (INIS)

    Beylot, J.

    1979-01-01

    Replacing the former highly enriched (93%) U-Al fuel by low enrichment (7%) oxide has brought about some changes in the experimental conditions for irradiations. The advantages for the experiments placed right in the lattice are shown to be a great improvement in the neutron spectrum (fast/thermal) and a very significant reduction in heating due to gamma radiation. In the case of the peripherally placed experiments there is an increase in the number of high thermal flux sites. In all cases, there is found to be an increase in the duration of the irradiation cycle, between two partial reloadings, permitted by the significant amount of 235 U tied up in the loading. The drawbacks observed are reduced thermo-hydraulic performance of the new fuel elements that does not allow working with a core of a size under a 7x7 configuration, increased surveillance of the kind of experiments placed in the lattice to avoid excessive power rises on the neighbouring fuel elements and moderate reduction in the level of thermal neutron fluxes in the peripheral irradiation sites [fr

  3. Non-standard constraints within In-Core Fuel Management

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Torres, C.; Marrote, G.N.; Ruiz U, V.

    2004-01-01

    Recent advancements in the area of nuclear fuel management optimization have been considerable and widespread. Therefore, it is not surprising that the design of today's nuclear fuel reloads can be a highly automated process that is often accompanied by sophisticated optimization software and graphical user interfaces to assist core designers. Most typically, among other objectives, optimization software seeks to maximize the energy efficiency of a fuel cycle while satisfying a variety of safety, operational, and regulatory constraints. Concurrently, the general industry trend continues to be one of pursuing higher generating capacity (i.e., power up-rates) alongside cycle length extensions. As these increasingly invaluable software tools and ambitious performance goals are pursued in unison, more aggressive core designs ultimately emerge that effectively minimize the margins to limits and, in some cases, may turn out less forgiving or accommodating to changes in underlying key assumptions. The purpose of this article is to highlight a few 'non-standard', though common constraints that can affect a BWR core design but which are often difficult, if not impossible, to implement into an automated setting. In a way, this article indirectly emphasizes the unique and irreplaceable role of the experienced designer in light of 'real life' situations. (Author)

  4. Pressurised water reactor fuel management using PANTHER

    International Nuclear Information System (INIS)

    Parks, G.T.; Knight, M.P.

    1996-01-01

    This paper describes the integration of Nuclear Electric's reactor physics code PANTHER with an automatic optimisation procedure designed to search for optimal PWR reload cores and assesses its performance. (Author)

  5. Affective Norms for 4900 Polish Words Reload (ANPW_R): Assessments for Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability and, Age of Acquisition

    Science.gov (United States)

    Imbir, Kamil K.

    2016-01-01

    In studies that combine understanding of emotions and language, there is growing demand for good-quality experimental materials. To meet this expectation, a large number of 4905 Polish words was assessed by 400 participants in order to provide a well-established research method for everyone interested in emotional word processing. The Affective Norms for Polish Words Reloaded (ANPW_R) is designed as an extension to the previously introduced the ANPW dataset and provides assessments for eight different affective and psycholinguistic measures of Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability, and subjective Age of Acquisition. The ANPW_R is now the largest available dataset of affective words for Polish, including affective scores that have not been measured in any other dataset (concreteness and age of acquisition scales). Additionally, the ANPW_R allows for testing hypotheses concerning dual-mind models of emotion and activation (origin and subjective significance scales). Participants in the current study assessed all 4905 words in the list within 1 week, at their own pace in home sessions, using eight different Self-assessment Manikin (SAM) scales. Each measured dimension was evaluated by 25 women and 25 men. The ANPW_R norms appeared to be reliable in split-half estimation and congruent with previous normative studies in Polish. The quadratic relation between valence and arousal was found to be in line with previous findings. In addition, nine other relations appeared to be better described by quadratic instead of linear function. The ANPW_R provides well-established research materials for use in psycholinguistic and affective studies in Polish-speaking samples. PMID:27486423

  6. Affective Norms for 4900 Polish Words Reload (ANPW_R:Assessments for Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability and Age of Acquisition.

    Directory of Open Access Journals (Sweden)

    Kamil Konrad Imbir

    2016-07-01

    Full Text Available In studies that combine understanding of emotions and language, there is growing demand for good-quality experimental materials. To meet this expectation, a large number of 4905 Polish words was assessed by 400 participants in order to provide a well-established research methodfor everyone interested in emotional word processing. The Affective Norms for Polish Words Reloaded (ANPW_R is designed as an extension to the previously introduced the ANPW dataset and provides assessments for eight different affective and psycholinguistic measures of Valence, Arousal, Dominance, Origin, Significance, Concreteness, Imageability and subjective Age of Acquisition. The ANPW_R is now the largest available dataset of affective words for Polish, including affective scores that have not been measured in any other dataset (concreteness and age of acquisition scales. Additionally, the ANPW_R allows for testing hypotheses concerning dual-mind models of emotion and activation (origin and subjective significance scales. Participants in the current study assessed all 4,905 words in the list within one week, at their own pace in home sessions, using eight different Self-assessment Manikin (SAM scales. Each measured dimension was evaluated by 25 women and 25 men. The ANPW_R norms appeared to be reliable in split-half estimation and congruent with previous normative studies in Polish. The quadratic relation between valence and arousal was found to be in line with previous findings. In addition, nine other relations appeared to be better described by quadratic instead of linear function. The ANPW_R provides well-established research materials for use in psycholinguistic and affective studies in Polish-speaking samples.

  7. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool

    International Nuclear Information System (INIS)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C.

    2014-10-01

    The fuel assemblies storage in the nuclear power plant of Laguna Verde (NPP-L V) represents a crucial aspect, due to the generated dose by the decay heat of the present radio-nuclides in the assemblies retired of the reactor core, after their useful life. These spent assemblies are located inside the spent fuel pool (SFP), in the level 49.90 m in the Reload Floor of the Reactor building of NPP-L V. This leads to the protection at personnel applying the ALARA (As Low As Reasonably Achievable) criteria, fulfilling the established dose criteria by the Regulator Body the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). Considering the loss scenario of the cooling system of the SFP, in which the SFP water vaporizes, is important to know the water level in which the limit of effective dose equivalent is fulfilled for the personnel. Also, is important for the instrumentation of the SFP, for the useful life of the same instruments. In this work is obtained the exposition rapidity corresponding to different water levels of SFP in the Reload Floor of NPP-L V, to identify the minimum level of water where the limit of effective dose equivalent is fulfilled of 25 rem s to the personnel, established in the Article 48 of the General Regulation of Radiological Safety of CNSNS and the Chapter 50 Section 67 of the 10-Cfr of Nuclear Regulatory Commission in USA. The water level is also identified where the exposition rapidity is of 15 m R/hr, being the value of the set point of the area radiation monitor D21-Re-N003-1, located to 125 cm over the level 49.90 meters of the Reload Floor of NPP-L V. (Author)

  8. Arrangement and statistics of storage containers of spent fuel for assemblies of the SFP of NPP-L V, Unit 1

    International Nuclear Information System (INIS)

    Mijangos D, Z. E.; Vargas A, A. F.; Amador C, C.

    2014-10-01

    This work presents the determination of assemblies of the spent fuel pool (SFP) of the nuclear power plant of Laguna Verde (NPP-L V) which are candidates to be assigned to storage containers of independent spent fuel, with the objective of liberating decay heat and to have more space in the SFP, for the store of retired assemblies of the reactors in future reloads of NPP-L V, besides that the removed assemblies of the SFP should be stored in specific containers to guarantee the physical safety of them, as well as the radiological protection to the population and the environment. The design of the containers considered in this work is to store a maximum of 69 assemblies; it has a thermal capacity of 26 kilowatts and allows storing assemblies with a minimum of 5 years of have been extracted of the reactor core. Is considered that in 2016 start the storage of the spent assemblies on the containers, the candidates assemblies to store cover from the first reload in 1991, until the assemblies deposited in the SFP in the 14 reload in 2010; therefore in 2016, such assemblies will have fulfilled with the criteria of 5 years of have been removed of the Reactor, also the 69 assemblies assigned to each container will have a resulting decay heat that does not exceed the thermal capacity of the container, but that in great percentage approximates to the same one, and this way to take full advantage of their storage capacity and thermal capacity for each container. This work also contains the arrangement to accommodate the assemblies in the containers; such arrangement is constituted by areas according to the decay heat of each assembly. (Author)

  9. LPG fuel

    International Nuclear Information System (INIS)

    Dagnas, F.X.; Jeuland, N.; Fouquet, J.P.; Lauraire, S.; Coroller, P.

    2005-01-01

    LPG fuel has become frequently used through a distribution network with 2 000 service stations over the French territory. LPG fuel ranks number 3 world-wide given that it can be used on individual vehicles, professional fleets, or public transport. What is the environmental benefit of LPG fuel? What is the technology used for these engines? What is the current regulation? Government commitment and dedication on support to promote LPG fuel? Car makers projects? Actions to favour the use of LPG fuel? This article gathers 5 presentations about this topic given at the gas conference

  10. Experience in the use of low concentration gadolinia as a PWR fuel burnable absorber

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Segovia, M.A.

    2001-01-01

    A description is provided of the low concentration gad design being used in the Spanish 3-loop 17 x 17 fueled PWR's. This design uses a relatively small number of high concentration gadolinia fuel rods (6 and 8 w/o Gd 2 O 3 ) with a large number of low concentration gad rods (2 w/o Gd 2 O 3 ). The 2 w/o gad rods substitute, in part, the high concentration gad rods, thereby helping reduce the end of cycle reactivity penalty from the residual absorption in the gadolinium. The low concentration gad design is advantageous for long cycles (18+ months) and plant up-rating scenarios in that the soluble boron concentration increases that would otherwise result for these situations are avoided. These boron concentration increases could have potentially adverse effects on the plant, since the moderator temperature coefficient (MTC) is made less negative, the effectiveness of the boron shutdown safety systems is reduced, and the safety margins are eroded for some accidents, such as for boron dilution events. These increases in the boron concentration would also require the plant to operate at higher lithium (Li) concentrations in the coolant in order to maintain the pH level at the desired value. Operation at the higher Li concentrations is undesirable because of the concerns over the potential impact on the fuel assembly material performance (e.g., crud and corrosion). This paper also reviews the APA (Alpha/Phoenix-P/ANC) nuclear design code system performance for the low concentration gad design. The design system performance for the reload cores that have or are employing this design has been completely satisfactory. The performance and accuracy of the nuclear design methodology is found to be as good for this design as for the reload cores that use exclusively high gad concentrations, or those that use WABA's - the discrete burnable absorber (BA) used prior to its substitution for gadolinium. (authors)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  12. Fuel Services

    International Nuclear Information System (INIS)

    Silberstein, A.

    1982-09-01

    FRAGEMA has developed most types of inspection equipments to work on irradiated fuel assemblies and on single fuel rods during reactor outages with an efficiency compatible with the utilities operating priorities. In order to illustrate this statement, two specific examples of inspection equipments are shortly described: the on-site removable fuel rod assembly examination stand, and the fuel assembly multiple examination device. FRAGEMA has developed techniques for the identifiction of the leaking fuel rods in the fuel assembly and the tooling necessary to perform the replacement of the faulted element. These examples of methods, techniques and equipments described and the experience accumulated through their use allow FRAGEMA to qualify for offering the supply of the corresponding software, hardware or both whenever an accurate understanding of the fuel behaviour is necessary and whenever direct intervention on the assembly and associated components is necessary due to safety, operating or economical reasons

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Watanabe, Shoichi; Hirano, Yasushi.

    1998-01-01

    A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

  14. The database system for the management of technical documentations of PWR fuel design project using CD-ROM

    International Nuclear Information System (INIS)

    Park, Bong Sik; Lee, Won Jae; Ryu, Jae Kwon; Jo, In Hang; Chang, Jong Hwa.

    1996-12-01

    In this report, the database system developed for the management of technical documentation of PWR fuel design project using CD-ROM (compact disk - read only memory) is described. The database system, KIRDOCM (KAERI Initial and Reload Fuel project technical documentation management), is developed and installed on PC using Visual Foxpro 3.0. Descriptions are focused on the user interface of the KIRDOCM. Introduction addresses the background and concept of the development. The main chapter describes the user requirements, the analysis of computing environment, the design of KIRDOCM, the implementation of the KIRDOCM, user's manual of KIRDOCM and the maintenance of the KIRDOCM for future improvement. The implementation of KIRDOCM system provides the efficiency in the management, maintenance and indexing of the technical documents. And, it is expected that KIRDOCM may be a good reference in applying Visual Foxpro for the development of information management system. (author). 2 tabs., 13 figs., 8 refs

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Matsuzuka, Ryuji.

    1976-01-01

    Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

  16. Fuel assemblies

    International Nuclear Information System (INIS)

    Nagano, Mamoru; Yoshioka, Ritsuo

    1983-01-01

    Purpose: To effectively utilize nuclear fuels by increasing the reactivity of a fuel assembly and reduce the concentration at the central region thereof upon completion of the burning. Constitution: A fuel assembly is bisected into a central region and a peripheral region by disposing an inner channel box within a channel box. The flow rate of coolants passing through the central region is made greater than that in the peripheral region. The concentration of uranium 235 of the fuel rods in the central region is made higher. In such a structure, since the moderating effect in the central region is improved, the reactivity of the fuel assembly is increased and the uranium concentration in the central region upon completion of the burning can be reduced, fuel economy and effective utilization of uranium can be attained. (Kamimura, M.)

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  18. Fuel assembly

    International Nuclear Information System (INIS)

    Yokota, Tokunobu.

    1990-01-01

    A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

  19. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  20. Pre-conceptual core design of SCWR with annular fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Chuanqi [Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Cao, Liangzhi, E-mail: caolz@mail.xjtu.edu.cn [Key Laboratory of Thermo-Fluid Science and Engineering of MOE, School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Wu, Hongchun; Zheng, Youqi [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China)

    2014-02-15

    Highlights: • Annular fuel with both internal and external cooling is used in supercritical light water reactor (SCWR). • The geometry of the annular fuel has been optimized to achieve better performance for the SCWR. • Based on the annular fuel assembly, an equilibrium core has been designed. • The results show that the equilibrium core has satisfied all the objectives and design criteria. - Abstract: The new design of supercritical light water reactor was proposed using annular fuel assemblies. Annular fuel consists of several concentric rings. Feed water flows through the center and outside of the fuel to give both internal and external cooling. Thanks to this feature, the fuel center temperature and the cladding temperature can be reduced and high power density can be achieved. The water flowing through the center also provides moderation, so there is no need for extra water rods in the assembly. The power distribution can be easily flattened by use of this design. The geometry of the annular fuel has been optimized to achieve better performance for the SCWR. There are 19 fuel pins in an assembly. Burnable poison is utilized to reduce the initial excess reactivity. The fuel reloading pattern and water flow scheme were optimized to achieve more uniform power distribution and lower cladding temperature. An equilibrium core has been designed and analyzed using three dimensional neutronics and thermal-hydraulics coupling calculations. The void reactivity, Doppler coefficient and cold shut down margin were calculated for safety consideration. The present results show that this concept is a promising design for the SCWR.

  1. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    International Nuclear Information System (INIS)

    Chodak, P. III

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO 2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239 Pu and ≥90% total Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products

  2. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Chodak, III, Paul [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1996-05-01

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO2 assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the 239Pu and ≥90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  3. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  4. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  5. Aims of failed fuel detection and substantiation of radiation safety at implementation of new kinds of nuclear fuel and fuel cycles on NPP with WWER

    International Nuclear Information System (INIS)

    Miglo, V.; Luzanova, L.

    2011-01-01

    Limiting of number of leaking fuel rods in a core during reactor operation in the analyses which are carried out for a substantiation of radiating safety for NPP with WWER as well as problems and possibilities of FFD at implementation of new kinds of fuel and fuel cycles are the main topics discussed in this paper. Available experience of designing of the NPP with WWER shows, that for ensuring of implementation of the RS criteria regarding limiting radioactive emissions from the NPP and doses of an irradiation of the population living near to NPP, it is required to regulate more rigidly number of failed fuel rods in comparison with requirements of Rules of nuclear safety NP-082-07. The reason of it is necessity to consider a technical condition of all safety barriers on a path of radioactive FP extension in a complex, first and foremost of uncontrolled leakage of the primary coolant to the NPP premises and efficiency of filters of ventilating systems, and also spike-effect on activity of isotopes of iodine after a power unit shutdown for fuel reloading and openings of a cover of a reactor. Depending on the project of NPP, parameters of fuel loading, a place of placing of the NPP and other factors the limit level of activity of isotopes of iodine in the primary coolant will be reached at various number of leaking fuel rods which can be unequal for various power units and the NPP with WWER, constructed on one design. The quantity of leaking fuel rods at which the design limit on FP-activity in the primary coolant of operating reactor is reached, can be essential below an operational limit on number of failed fuel rods established by Rules of nuclear safety. However the reached quality of fabrication of the WWER fuel rods providing their high reliability (the probability of fuel rod failure in the course of one operation year is not higher than 10 -5 ) as well as due to the levels of the WWER fuel rod depressurization actually attainable in the normal conditions of

  6. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Bassler, E.A.; Huckestein, E.A.; Salton, R.B.; Tower, S.N.

    1988-01-01

    A fuel assembly adapted for use with a pressurized water nuclear reactor having capabilities for fluid moderator spectral shift control is described comprising: parallel arranged elongated nuclear fuel elements; means for providing for axial support of the fuel elements and for arranging the fuel elements in a spaced array; thimbles interspersed among the fuel elements adapted for insertion of a rod control cluster therewithin; means for structurally joining the fuel elements and the guide thimbles; fluid moderator control means for providing a volume of low neutron absorbing fluid within the fuel assembly and for removing a substantially equivalent volume of reactor coolant water therefrom, a first flow manifold at one end of the fuel assembly sealingly connected to a first end of the moderator control tubes whereby the first ends are commonly flow connected; and a second flow manifold, having an inlet passage and an outlet passage therein, sealingly connected to a second end of the moderator control tubes at a second end of the fuel assembly

  9. Influence of FGR complexity modelling on the practical results in gas pressure calculation of selected fuel elements from Dukovany NPP

    International Nuclear Information System (INIS)

    Lahodova, M.

    2001-01-01

    A modernization fuel system and advanced fuel for operation up to the high burnup are used in present time in Dukovany NPP. Reloading of the cores are evaluated using computer codes for thermomechanical behavior of the most loaded fuel rods. The paper presents results of parametric calculations performed by the NRI Rez integral code PIN, version 2000 (PIN2k) to assess influence of fission gas release modelling complexity on achieved results. The representative Dukovany NPP fuel rod irradiation history data are used and two cases of fuel parameter variables (soft and hard) are chosen for the comparison. Involved FGR models where the GASREL diffusion model developed in the NRI Rez plc and standard Weisman model that is recommended in the previous version of the PIN integral code. FGR calculation by PIN2k with GASREL model represents more realistic results than standard Weisman's model. Results for linear power, fuel centre temperature, FGR and gas pressure versus burnup are given for two fuel rods

  10. Considerations for a national program on spent fuel management

    International Nuclear Information System (INIS)

    Lopez-Perez, B.; Melches-Serrano, C.

    1980-01-01

    The spent fuel discharged from the two LWR's that are in operation (Zorita, 160 MW PWR, and Santa Maria de Garona, 460 MW BWR) is being reprocessed under contracts with BNFL; these contracts will expire in the next few years. The fuel discharged from Vandelos (50 MW GCR) is being reprocessed by Cogema under a long-term contract. No new reprocessing contracts for LWR's in operation, under construction, or planned have been signed or are being considered for the near future. The plutonium and the residual uranium contained in LWR spent fuel are considered important potential energy resources. They are especially valuable for countries such as Spain, which is short of energy resources, and they might be used in the future in fast breeder or thermal reactors. This is the reason that, until reprocessing is justified and appropriate solutions to make reprocessing available are developed, Spain has decided to build the appropriate capacity for the temporary storage of spent fuel. The capacity is being achieved, on short term, by the extension of AR storage capacity. It is being achieved, at medium or longer term, by the construction of centralized AFR facilities to serve all Spanish nuclear power plants. Spanish utilities are undertaking the expansion of reactor storage capacities, using densified racks, to increment capacity to at least 8 to 10 reloads, in addition to full core discharge capacity. Spain has the time and the financial and technical resources to implement a national solution for spent fuel storage. Financial strategy, technology choice, and licensing considerations are under examination in order to make a decision for medium- and long-term storage alternatives

  11. User's guide for the REBUS-3 fuel cycle analysis capability

    Energy Technology Data Exchange (ETDEWEB)

    Toppel, B.J.

    1983-03-01

    REBUS-3 is a system of programs designed for the fuel-cycle analysis of fast reactors. This new capability is an extension and refinement of the REBUS-3 code system and complies with the standard code practices and interface dataset specifications of the Committee on Computer Code Coordination (CCCC). The new code is hence divorced from the earlier ARC System. In addition, the coding has been designed to enhance code exportability. Major new capabilities not available in the REBUS-2 code system include a search on burn cycle time to achieve a specified value for the multiplication constant at the end of the burn step; a general non-repetitive fuel-management capability including temporary out-of-core fuel storage, loading of fresh fuel, and subsequent retrieval and reloading of fuel; significantly expanded user input checking; expanded output edits; provision of prestored burnup chains to simplify user input; option of fixed-or free-field BCD input formats; and, choice of finite difference, nodal or spatial flux-synthesis neutronics in one-, two-, or three-dimensions.

  12. SunFast: A sun workstation based, fuel analysis scoping tool for pressurized water reactors

    International Nuclear Information System (INIS)

    Bohnhoff, W.J.

    1991-05-01

    The objective of this research was to develop a fuel cycle scoping program for light water reactors and implement the program on a workstation class computer. Nuclear fuel management problems are quite formidable due to the many fuel arrangement options available. Therefore, an engineer must perform multigroup diffusion calculations for a variety of different strategies in order to determine an optimum core reload. Standard fine mesh finite difference codes result in a considerable computational cost. A better approach is to build upon the proven reliability of currently available mainframe computer programs, and improve the engineering efficiency by taking advantage of the most useful characteristic of workstations: enhanced man/machine interaction. This dissertation contains a description of the methods and a user's guide for the interactive fuel cycle scoping program, SunFast. SunFast provides computational speed and accuracy of solution along with a synergetic coupling between the user and the machine. It should prove to be a valuable tool when extensive sets of similar calculations must be done at a low cost as is the case for assessing fuel management strategies. 40 refs

  13. Design of reactor internals in larger high-temperature reactors with spherical fuel elements

    International Nuclear Information System (INIS)

    Elter, C.

    1981-01-01

    In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL) [de

  14. Fuel element

    International Nuclear Information System (INIS)

    Kennedy, S.T.

    1982-01-01

    A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

  15. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

  16. Fuel cells

    NARCIS (Netherlands)

    Veen, van J.A.R.; Janssen, F.J.J.G.; Santen, van R.A.

    1999-01-01

    The principles and present-day embodiments of fuel cells are discussed. Nearly all cells are hydrogen/oxygen ones, where the hydrogen fuel is usually obtained on-site from the reforming of methane or methanol. There exists a tension between the promise of high efficiency in the conversion of

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  18. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  19. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  20. Fuel and Core Design Verification for Extended Power Up-rate in Ringhals Unit 3

    International Nuclear Information System (INIS)

    Gabrielsson, Petter; Stepniewski, Marek; Almberger, Jan

    2006-01-01

    Vattenfall's Westinghouse 3-loop PWR Ringhals 3 at the western coast of Sweden is scheduled for an extended power up-rate from 2783 to 3160 MWt in 2007, in the frame of the so called GREAT-project. The project will realize an up-rating initially planned and analysed back in 1995, but with a number of significant improvements outlined in this paper. For the licensing of the up-rated power level, a complete revision of the safety analyses, radiological analyses and systems verifications in FSAR is being performed by Westinghouse Electrics Belgium. The work is performed in close cooperation with Vattenfall in the areas of core calculations and input data. For more than a decade, Vattenfall has performed all core design and reload safety evaluations (RSE) for Ringhals, independent of fuel vendors and safety analysts. In GREAT all core parameters in the safety analysis checklist (SAC) used for the safety analyses are determined based upon a set of nine reference loading patterns designed by Vattenfall covering a wide range of fuel and core designs and extreme cycle-to-cycle variations. To facilitate the calculation of SAC parameters Westinghouse has provided a Reload Safety Evaluation Procedure report (RSEP) with detailed specifications for the calculation of all core parameters used in the analyses. The procedure has been automatized by Vattenfall in a set of scripts executing 3D core simulator calculations and extracting the key results. The same tools will be used in Vattenfall's future RSE for Ringhals 3. This approach is taken to obtain consistency between core designs and core calculations for the safety analyses and the cycle specific calculations, to minimize the risk for future violations of the safety analyses. (authors)

  1. Technology of the production of breeder fuel elements

    International Nuclear Information System (INIS)

    Funke, P.

    1976-01-01

    A survey is presented of the fabrication of oxide and carbide fuels and of the fuel rod for fast breeders (KNK, SNR-300). The advantages of the chosen methods are explained. The main points of development concerning the oxide fuel rod are gone into. The process sequence for plutonium oxide and plutonium carbide processing is presented in a flow chart. (HR) [de

  2. Diuretics in primary hypertension - Reloaded.

    Science.gov (United States)

    Mishra, Sundeep

    Diuretics have long been cherished as drugs of choice for uncomplicated primary hypertension. Robust mortality and morbidity data is available for diuretics to back this strategy. Off-late the interest for diuretics has waned off perhaps due to availability of more effective drugs but more likely due to perceived lack of tolerance and side-effect profile of high-dose of diuretics required for mortality benefit. Low-dose diuretics particularly thiazide diuretics are safer but lack the mortality benefit shown by high-dose. However, indapamide and low dose chlorthalidone have fewer side-effects but continue to provide mortality benefit. Copyright © 2016. Published by Elsevier B.V.

  3. Data mining in the study of nuclear fuel cells; Mineria de datos en el estudio de celdas de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Medina P, J. A. [Universidad Autonoma de Campeche, Av. Agustin Melgar s/n, Col. Buenavista, 24039 San Francisco de Campeche, Campeche (Mexico); Ortiz S, J. J.; Castillo, A.; Montes T, J. L.; Perusquia, R., E-mail: j.angel.mp@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    In this paper is presented a study of data mining application in the analysis of fuel cells and their performance within a nuclear boiling water reactor. A decision tree was used to fulfill questions of the type If (condition) and Then (conclusion) to classify if the fuel cells will have good performance. The performance is measured by compliance or not of the cold shutdown margin, the rate of linear heat generation and the average heat generation in a plane of the reactor. It is assumed that the fuel cells are simulated in the reactor under a fuel reload and rod control patterns pre designed. 18125 fuel cells were simulated according to a steady-state calculation. The decision tree works on a target variable which is one of the three mentioned before. To analyze this objective, the decision tree works with a set of attribute variables. In this case, the attributes are characteristics of the cell as number of gadolinium rods, rods number with certain uranium enrichment mixed with a concentration of gadolinium, etc. The found model was able to predict the execution or not of the shutdown margin with a precision of around 95%. However, the other two variables showed lower percentages due to few learning cases of the model in which these variables were or were not achieved. Even with this inconvenience, the model is quite reliable and can be used in way coupled in optimization systems of fuel cells. (Author)

  4. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  5. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  6. Method to generate the first design of the reload pattern to be used with the Presto-B code in the simulation of the CNLV U-1 reactor; Metodo para generar el primer diseno del patron de recarga a ser utilizado con el codigo Presto-B, en la simulacion del reactor de la CNLV U-1

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Cortes C, C.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1992-08-15

    This guide is applied for the reload pattern's formation for mirror symmetry of a core room and in accordance with the Control Cell core technique (of the english Control Cell Core - CCC) for the PRESTO-B code. (Author)

  7. New fuel vault criticality analysis at Chinshan nuclear power station with new approaches to improve the storage flexibility

    International Nuclear Information System (INIS)

    Huang, P. H.

    2010-10-01

    The Chinshan new fuel vault (NFV) consists of 13 fuel storage racks, each rack may store 10 fuel assemblies. Prior to 2008, the NFV had never been used and the practice by the Taiwan Power Company (TPC) was to temporarily store the fuel assemblies in the shipping containers after received, until the inspection work was performed shortly before the outage, and then assemblies were loaded directly into the spent fuel pool (SFP). Starting from 2009, this practice has been revised since the new fuel contract would only supply a small amount of containers for storage, and the SFP would lose full-core-off load capability soon; therefore, use of NFV to store fuel assemblies following inspection becomes extremely crucial. The original Chinshan NFV criticality analysis was performed for the initial fuel design. Although many new fuel designs had been used (e.g., Atrium-10 reported in PBNC-14), no reanalysis had been performed because it was not anticipated that NFV would be used. Therefore, TPC requested the vendor to perform the analysis for Atrium-10. Originally, the vendor estimated that number of assemblies allowed to be stored would be limited severely to about 60. To enhance storage flexibility, Tpc proposed some new approaches: 1) All assemblies are assumed in vendor's standard method to contain a single limiting lattice for entire fuel length, it is suggested that axially zoned limiting lattices be selected based on characteristics of reloads to be delivered, and this significantly improves flexibility. 2) The maximum k-effective equation used by vendor was corrected (manufacturing tolerances were conservatively mistreated). Also, the vendor typically used 0.95 k-effective as the criterion, it is suggested that NUREG-0800 requirement (≤0.98 for optimum moderation conditions) be applied. After several iterations, all the 130 locations are allowed to store fuel. The analysis report has been approved by the authority in June 2008. (Author)

  8. New fuel vault criticality analysis at Chinshan nuclear power station with new approaches to improve the storage flexibility

    Energy Technology Data Exchange (ETDEWEB)

    Huang, P. H., E-mail: u808966@taipower.com.t [Taiwan Power Company, Department of Nuclear Generation, 242 Roosevelt Rd., Sec. 3, Taipei, Taiwan (China)

    2010-10-15

    The Chinshan new fuel vault (NFV) consists of 13 fuel storage racks, each rack may store 10 fuel assemblies. Prior to 2008, the NFV had never been used and the practice by the Taiwan Power Company (TPC) was to temporarily store the fuel assemblies in the shipping containers after received, until the inspection work was performed shortly before the outage, and then assemblies were loaded directly into the spent fuel pool (SFP). Starting from 2009, this practice has been revised since the new fuel contract would only supply a small amount of containers for storage, and the SFP would lose full-core-off load capability soon; therefore, use of NFV to store fuel assemblies following inspection becomes extremely crucial. The original Chinshan NFV criticality analysis was performed for the initial fuel design. Although many new fuel designs had been used (e.g., Atrium-10 reported in PBNC-14), no reanalysis had been performed because it was not anticipated that NFV would be used. Therefore, TPC requested the vendor to perform the analysis for Atrium-10. Originally, the vendor estimated that number of assemblies allowed to be stored would be limited severely to about 60. To enhance storage flexibility, Tpc proposed some new approaches: 1) All assemblies are assumed in vendor's standard method to contain a single limiting lattice for entire fuel length, it is suggested that axially zoned limiting lattices be selected based on characteristics of reloads to be delivered, and this significantly improves flexibility. 2) The maximum k-effective equation used by vendor was corrected (manufacturing tolerances were conservatively mistreated). Also, the vendor typically used 0.95 k-effective as the criterion, it is suggested that NUREG-0800 requirement ({<=}0.98 for optimum moderation conditions) be applied. After several iterations, all the 130 locations are allowed to store fuel. The analysis report has been approved by the authority in June 2008. (Author)

  9. Fuel behaviour

    International Nuclear Information System (INIS)

    Fodor, M.; Matus, L.; Vigassy, J.

    1987-11-01

    A short summary of the main critical points in fuel performance of nuclear power reactors from chemical and mechanical point of view is given. A schedule for a limited research program is included. (author) 17 refs

  10. Fuel cells

    International Nuclear Information System (INIS)

    Niederdoeckl, J.

    2001-01-01

    Europe has at present big hopes on the fuel cells technology, in comparison with other energy conversion technologies, this technology has important advantages, for example: high efficiency, very low pollution and parallel use of electric and thermal energy. Preliminary works for fuel cells developing and its commercial exploitation are at full speed; until now the European Union has invested approx. 1.7 billion Schillings, 60 relevant projects are being executed. The Austrian industry is interested in applying this technique to drives, thermal power stations and the miniature fuel cells as replacement of batteries in electronic products (Notebooks, mobile telephones, etc.). A general description of the historic development of fuel cells including the main types is given as well as what is the situation in Austria. (nevyjel)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

    1995-01-01

    Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

  12. Fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1976-01-01

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

  13. A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

    International Nuclear Information System (INIS)

    Polley, M.V.; Evans, H.E.

    1993-08-01

    High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion

  14. Nuclear fuel management and transients analysis in Laguna Verde nuclear power plant

    International Nuclear Information System (INIS)

    De Loera De Haro, M.A.; Alvarez Gasca, J.

    1991-01-01

    Nuclear fuel management transient analysis are the set of activities which determine the load and reload of nuclear fuel inside the reactor, with the aim of getting the maximum performance in fuel burn up and heat remotion, without have an effect in the station safety. Nuclear fuel management and transient analysis has its basis on high precision quantitative analysis methodologies by means of simulation of nuclear and physical phenomena occurring both in normal and abnormal operation of nuclear power plants. On account of complexity of simulations and the required precision, those are carry out using codes type 'best estimate'. For the use of this tools it is necessary a deep knowledge of simulated nuclear and physical phenomena, as well as the used mathematical models and the numerical methods used. If different, the simulation results will be notably different actual processes owing to the use of models out of validity range, or incorrect calculations in the input parameters. On account of complexity of simulations and the required precision, those are carry out using codes type 'best estimate'. For the use of this tools it is necessary a deep knowledge of simulated nuclear and physical phenomena, as well as the used mathematical models and the numerical methods used. If different, the simulation results will be notably different actual processes owing to the use of models out of validity range, or incorrect calculations in the input parameters

  15. High conversion Th-U{sup 233} fuel assembly for current generation of PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Baldova, D.; Fridman, E. [Reactor Safety Div., Helmholtz-Zentrum Dresden-Rossendorf, POB 510119, Dresden, 01314 (Germany)

    2012-07-01

    This paper presents a preliminary design of a high conversion Th-U{sup 233} fuel assembly applicable for current generation of Pressurized Water Reactor (PWRs). The considered fuel assembly has a typical 17 x 17 PWR lattice. However in order to increase the conversion of Th{sup 232} to U{sup 233}, the assembly was subdivided into the two regions called seed and blanket. The central seed region has a higher than blanket U{sup 233} content and acts as a neutron source for the peripheral blanket region. The latest acts as a U{sup 233} breeder. While the seed fuel pins have a standard dimensions the blanket fuel radius was increased in order to reduce the moderation and to facilitate the resonance neutron absorption in blanket Th{sup 232}. The U{sup 233} content in the seed and blanket regions was optimized to achieve maximal initial to discharged fissile inventory ratio (FIR) taking into account the target fuel cycle length of 12 months with 3-batch reloading scheme. In this study the neutronic calculations were performed on the fuel assembly level using Helios deterministic lattice transport code. The fuel cycle length and the core k{sub eff} were estimated by applying the Non Linear Reactivity Model. The applicability of the HELIOS code for the analysis of the Th-based high conversion designs was confirmed with the help of continuous-energy Monte-Carlo code SERPENT. The results of optimization studies show that for the heterogeneous seed and blanket (SB) fuel assembly the FIR of about 0.95 can be achieved. (authors)

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  17. Statistical analysis in the design of nuclear fuel cells and training of a neural network to predict safety parameters for reactors BWR; Analisis estadistico en el diseno de celdas de combustible nuclear y entrenamiento de una red neuronal para predecir parametros de seguridad para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Jauregui Ch, V.

    2013-07-01

    In this work the obtained results for a statistical analysis are shown, with the purpose of studying the performance of the fuel lattice, taking into account the frequency of the pins that were used. For this objective, different statistical distributions were used; one approximately to normal, another type X{sup 2} but in an inverse form and a random distribution. Also, the prediction of some parameters of the nuclear reactor in a fuel reload was made through a neuronal network, which was trained. The statistical analysis was made using the parameters of the fuel lattice, which was generated through three heuristic techniques: Ant Colony Optimization System, Neuronal Networks and a hybrid among Scatter Search and Path Re linking. The behavior of the local power peak factor was revised in the fuel lattice with the use of different frequencies of enrichment uranium pines, using the three techniques mentioned before, in the same way the infinite multiplication factor of neutrons was analyzed (k..), to determine within what range this factor in the reactor is. Taking into account all the information, which was obtained through the statistical analysis, a neuronal network was trained; that will help to predict the behavior of some parameters of the nuclear reactor, considering a fixed fuel reload with their respective control rods pattern. In the same way, the quality of the training was evaluated using different fuel lattices. The neuronal network learned to predict the next parameters: Shutdown Margin (SDM), the pin burn peaks for two different fuel batches, Thermal Limits and the Effective Neutron Multiplication Factor (k{sup eff}). The results show that the fuel lattices in which the frequency, which the inverted form of the X{sup 2} distribution, was used revealed the best values of local power peak factor. Additionally it is shown that the performance of a fuel lattice could be enhanced controlling the frequency of the uranium enrichment rods and the variety of

  18. Shotgun protein sequencing.

    Energy Technology Data Exchange (ETDEWEB)

    Faulon, Jean-Loup Michel; Heffelfinger, Grant S.

    2009-06-01

    A novel experimental and computational technique based on multiple enzymatic digestion of a protein or protein mixture that reconstructs protein sequences from sequences of overlapping peptides is described in this SAND report. This approach, analogous to shotgun sequencing of DNA, is to be used to sequence alternative spliced proteins, to identify post-translational modifications, and to sequence genetically engineered proteins.

  19. Feasibility study of power reactor fuel elements factory development: I. Economical aspects

    International Nuclear Information System (INIS)

    Marwoto; Ratih-Langenati, R.R.; Susanti, P.

    1996-01-01

    For determining the feasibility study on manufacturing nuclear fuel element from economical aspect point of view, it necessary to fix its capacity which it was found from fuel element reloading requirement for nuclear power plat (PLTN). NEWJEC report which use as a base in this study that is possibly of a complex of NPP as big as 7200 MW in Muria region. If the capacity factor is 80 %, the reload requirement is therefore become from 120 to 142 tons uranium every year. So, its considered to fix the nominal capacity of a fabric for nuclear fuel element manufacturing as much as 200 tons-U per year with economical lifetimes of 20 years. NEWJEC data show, for manufacturing capacity of 200 tons-U per year with, plant have a fixed capital investment of US$ 43.9 million. With working capital as much as 15 % correspond to fixed capital investment (FCI); 10 % of interest rate; US$ 17 million of fixed cost; US$ 106.2/kg-U of variable production cost, its calculated that break even point/BEP is 50 % for price of nuclear fuel is US$ 350/kg-U without uranium cost. On this economic condition, it was found that the return on investment/ROI is 20.2 %; the internal rate of return/IRR is 11.2 % and the benefit cost ration/BCR is 1.22. For all of above, it was assumed that such nuclear fuel element manufacturing service will be operate in the year of 2012. Some of NEWJEC data have been revised, there were the value of FCI; cost of salary; the value in percent of working capital/WC; the cost of non-uranium materials and the price of product service are US$ 68 million; US$ 4.1 million; 30 %; US$ 100/kg-U and US$ 370/kg-U respectively, where the new data appear as higher than old date from NEWJEC, excluding the cost of salary. For all new economical data in the latest, we found that 45 %; 16.73 %; 11.8 % and 1.25 for BEB; IRR and BCR respectively

  20. Roles of programmable logic controllers in fuel reprocessing plants

    International Nuclear Information System (INIS)

    Mishra, Hrishikesh; Balakrishnan, V.P.; Pandya, G.J.

    1999-01-01

    Fuel charging facility is another application of Programmable Logic Controllers (PLC) in fuel reprocessing plants, that involves automatic operation of fuel cask dolly, charging motor, pneumatic doors, clutches, clamps, stepper motors and rod pushers in a pre-determined sequence. Block diagram of ACF system is given for underlining the scope of control and interlocks requirements involved for automation of the fuel charging system has been provided for the purpose at KARP Plant, Kalpakkam

  1. Transversal stiffness and beta-actin and alpha-actinin-4 content of the M. soleus fibers in the conditions of a 3-day reloading after 14-day gravitational unloading.

    Science.gov (United States)

    Ogneva, I V

    2011-01-01

    The aim of the work was to analyze the structural changes in different parts of the sarcolemma and contractile apparatus of muscle fibers by measuring their transversal stiffness by atomic force microscopy in a three-day reloading after a 14-day gravity disuse, which was carried out by hind-limbs suspension. The object of the study was the soleus muscle of the Wistar rat. It was shown that after 14 days of disuse, there was a reduction of transversal stiffness of all points of the sarcolemma and contractile apparatus. Readaptation for 3 days leads to complete recovery of the values of the transversal stiffness of the sarcolemma and to partial value recovery of the contractile apparatus. The changes in transversal stiffness of sarcolemma correlate with beta-actin and alpha-actinin-4 in membrane protein fractions.

  2. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  3. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  5. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  6. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  7. The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-01

    The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

  8. Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

    International Nuclear Information System (INIS)

    Hernandez G, J.

    2012-10-01

    Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

  9. Reliability assessment of the fueling machine of the CANDU reactor

    International Nuclear Information System (INIS)

    Al-Kusayer, T.A.

    1985-01-01

    Fueling of CANDU-reactors is carried out by two fueling machines, each serving one end of the reactor. The fueling machine becomes a part of the primary heat transport system during the refueling operations, and hence, some refueling machine malfunctions could result in a small scale-loss-of-coolant accident. Fueling machine failures and the failure sequences are discussed. The unavailability of the fueling machine is estimated by using fault tree analysis. The probability of mechanical failure of the fueling machine interface is estimated as 1.08 x 10 -5 . (orig.) [de

  10. Multimodal sequence learning.

    Science.gov (United States)

    Kemény, Ferenc; Meier, Beat

    2016-02-01

    While sequence learning research models complex phenomena, previous studies have mostly focused on unimodal sequences. The goal of the current experiment is to put implicit sequence learning into a multimodal context: to test whether it can operate across different modalities. We used the Task Sequence Learning paradigm to test whether sequence learning varies across modalities, and whether participants are able to learn multimodal sequences. Our results show that implicit sequence learning is very similar regardless of the source modality. However, the presence of correlated task and response sequences was required for learning to take place. The experiment provides new evidence for implicit sequence learning of abstract conceptual representations. In general, the results suggest that correlated sequences are necessary for implicit sequence learning to occur. Moreover, they show that elements from different modalities can be automatically integrated into one unitary multimodal sequence. Copyright © 2015 Elsevier B.V. All rights reserved.

  11. Sequence Read Archive (SRA)

    Data.gov (United States)

    U.S. Department of Health & Human Services — The Sequence Read Archive (SRA) stores raw sequencing data from the next generation of sequencing platforms including Roche 454 GS System®, Illumina Genome...

  12. Department of the Navy final environmental impact statement for a container system for the management of naval spent nuclear fuel

    International Nuclear Information System (INIS)

    1996-11-01

    This Final Environmental Impact Statement (EIS) addresses six general alternative systems for the loading, storage, transport, and possible disposal of naval spent nuclear fuel following examination. This EIS describes environmental impacts of (1) producing and implementing the container systems (including those impacts resulting from the addition of the capability to load the containers covered in this EIS in dry fuel handling facilities at Idaho National Engineering Laboratory (INEL)); (2) loading of naval spent nuclear fuel at the Expended Core Facility or at the Idaho Chemical Processing Plant with subsequent storage at INEL; (3) construction of a storage facility (such as a paved area) at alternative locations at INEL; and (4) loading of containers and their shipment to a geologic repository or to a centralized interim storage site outside the State of Idaho once one becomes available. As indicated in the EIS, the systems and facilities might also be used for handling low-level radiological waste categorized as special case waste. The Navy's preferred alternative for a container system for the management of naval spent fuel is a dual-purpose canister system. The primary benefits of a dual-purpose canister system are efficiencies in container manufacturing and fuel reloading operations, and potential reductions in radiation exposure

  13. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  14. Matlab enhanced multi-threaded tomography optimization sequence (MEMTOS)

    International Nuclear Information System (INIS)

    Lum, Edward S.; Pope, Chad L.

    2016-01-01

    Highlights: • Monte Carlo simulation of spent nuclear fuel assembly neutron computed tomography. • Optimized parallel calculations conducted from within the MATLAB environment. • Projection difference technique used to identify anomalies in spent nuclear fuel assemblies. - Abstract: One challenge associated with spent nuclear fuel assemblies is the lack of non-destructive analysis techniques to determine if fuel pins have been removed or replaced or if there are significant defects associated with fuel pins deep within a fuel assembly. Neutron computed tomography is a promising technique for addressing these qualitative issues. Monte Carlo simulation of spent nuclear fuel neutron computed tomography allows inexpensive process investigation and optimization. The main purpose of this work is to provide a fully automated advanced simulation framework for the analysis of spent nuclear fuel inspection using neutron computed tomography. The simulation framework, called Matlab Enhanced Multi-Threaded Tomography Optimization Sequence (MEMTOS) not only automates the simulation process, but also generates superior tomography image results. MEMTOS is written in the MATLAB scripting language and addresses file management, parallel Monte Carlo execution, results extraction, and tomography image generation. This paper describes the mathematical basis for neutron computed tomography, the Monte Carlo technique used to simulate neutron computed tomography, and the overall tomography simulation optimization algorithm. Sequence results presented include overall simulation speed enhancement, tomography and image results obtained for Experimental Breeder Reactor II spent fuel assemblies and light water reactor fuel assemblies. Optimization using a projection difference technique are also described.

  15. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

    2013-01-01

    strategic planning for fuel management such as shuffling and/or reloading schemes and its safe implementation.

  16. Method for compacting spent nuclear reactor fuel rods

    International Nuclear Information System (INIS)

    Wachter, W.J.

    1988-01-01

    In a nuclear reactor system which requires periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly is described. The method consists of: (1) removing the top end from the fuel rod assembly; (2) passing each of multiple fuel rod pulling elements in sequence through a fuel rod container and thence through respective consolidating passages in a fuel rod directing chamber; (3) engaging one of the pulling elements to the top end of each of the fuel rods; (4) drawing each of the pulling elements axially to draw the respective engaged fuel rods in one axial direction through the respective the passages in the chamber to thereby consolidate the fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within the fuel rod assembly; and (5) drawing all of the engaged fuel rods concurrently and substantially parallel to one another in the one axial direction into the fuel rod container while maintaining the compacted configuration whereby the fuel rods are aligned within the container in a fuel rod density of the the fuel rod assembly

  17. Fuels characterization studies. [jet fuels

    Science.gov (United States)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  18. Alternative Fuels Data Center: Ethanol Fueling Stations

    Science.gov (United States)

    ... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure fueling stations by location or along a route. Infrastructure Development Learn about ethanol fueling infrastructure; codes, standards, and safety; and ethanol equipment options. Maps & Data E85 Fueling Station

  19. Alternative Fuels Data Center: Biodiesel Fueling Stations

    Science.gov (United States)

    Locations Infrastructure Development Vehicles Laws & Incentives Biodiesel Fueling Stations Photo of a location or along a route. Infrastructure Development Learn about biodiesel fueling infrastructure codes Case Studies California Ramps Up Biofuels Infrastructure Green Fueling Station Powers Fleets in Upstate

  20. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Kurihara, Kunitoshi; Azekura, Kazuo.

    1992-01-01

    In a reactor core of a heavy water moderated light water cooled pressure tube type reactor, no sufficient effects have been obtained for the transfer width to a negative side of void reactivity change in a region of a great void coefficient. Then, a moderation region divided into upper and lower two regions is disposed at the central portion of a fuel assembly. Coolants flown into the lower region can be discharged to the cooling region from an opening disposed at the upper end portion of the lower region. Light water flows from the lower region of the moderator region to the cooling region of the reactor core upper portion, to lower the void coefficient. As a result, the reactivity performance at low void coefficient, i.e., a void reaction rate is transferred to the negative side. Thus, this flattens the power distribution in the fuel assembly, increases the thermal margin and enables rapid operaiton and control of the reactor core, as well as contributes to the increase of fuel burnup ratio and reduction of the fuel cycle cost. (N.H.)

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

    1998-01-01

    In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

  3. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  4. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  5. Analysis of dismantling possibility and unloading efforts of fuel assemblies from core of WWER

    International Nuclear Information System (INIS)

    Danilov, V.; Dobrov, V.; Semishkin, V.; Vasilchenko, I.

    2006-01-01

    The computation methods of optimal dismantling sequence of fuel assemblies (FA) from core of WWER after different operating periods and accident conditions are considered. The algorithms of fuel dismantling sequence are constructed both on the basis of analysis of mutual spacer grid overlaps of adjacent fuel assemblies and numerical structure analysis of efforts required for FA removal as FA heaving from the core. Computation results for core dismantling sequence after 3-year operating period and LB LOCA are presented in the paper

  6. Thorium fuel cycle management

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  7. Fast breeder fuel pin bundle tests in the KNK II-reactor

    International Nuclear Information System (INIS)

    Haefner, H.E.; Bojarsky, E.

    1986-11-01

    Three variants of ring elements with test bundles will be reported in this paper: In a first step a ring element was built with a permanently integrated test bundle (19 carbide pins of the Karlsruhe reference concept) while the proven fuel element components have been largely maintained. This irradiation will be completed in autumn 1986 after 380 full power days of operation. The central topic of this paper will be the technique of reloadable ring elements with replaceable test bundles. A first experiment, TOAST, is in preparation. For this experiment, above all the components of the fuel element head and foot had to be newly developed and tested. A special version of double-walled replaceable test bundles to be used in the TETRA temperature transient experiments will be briefly mentioned. It is envisaged in these experiments to vary in a defined manner the coolant flow at remotely assembled test bundles consisting of 19 KNK pins each having undergone a high burnup and to use a measuring and control plug placed on the test bundle so that a variety of fuel pin temperature programs can be realized. Finally, some additional aspects of bundle design will be indicated. (orig./GL) [de

  8. Repairing fuel for reinsertion

    International Nuclear Information System (INIS)

    Krukshenk, A.

    1986-01-01

    Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

  9. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

    Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

  10. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool; Determinacion de la rapidez de exposion en el nivel 49.90 del edificio del reactor por la disminucion en el nivel de agua de la alberca de combustible gastado

    Energy Technology Data Exchange (ETDEWEB)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C., E-mail: zoedelfin@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The fuel assemblies storage in the nuclear power plant of Laguna Verde (NPP-L V) represents a crucial aspect, due to the generated dose by the decay heat of the present radio-nuclides in the assemblies retired of the reactor core, after their useful life. These spent assemblies are located inside the spent fuel pool (SFP), in the level 49.90 m in the Reload Floor of the Reactor building of NPP-L V. This leads to the protection at personnel applying the ALARA (As Low As Reasonably Achievable) criteria, fulfilling the established dose criteria by the Regulator Body the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). Considering the loss scenario of the cooling system of the SFP, in which the SFP water vaporizes, is important to know the water level in which the limit of effective dose equivalent is fulfilled for the personnel. Also, is important for the instrumentation of the SFP, for the useful life of the same instruments. In this work is obtained the exposition rapidity corresponding to different water levels of SFP in the Reload Floor of NPP-L V, to identify the minimum level of water where the limit of effective dose equivalent is fulfilled of 25 rem s to the personnel, established in the Article 48 of the General Regulation of Radiological Safety of CNSNS and the Chapter 50 Section 67 of the 10-Cfr of Nuclear Regulatory Commission in USA. The water level is also identified where the exposition rapidity is of 15 m R/hr, being the value of the set point of the area radiation monitor D21-Re-N003-1, located to 125 cm over the level 49.90 meters of the Reload Floor of NPP-L V. (Author)

  11. Arrangement and statistics of storage containers of spent fuel for assemblies of the SFP of NPP-L V, Unit 1; Arreglo y estadistica de contenedores de almacenamiento de combustible gastado para los ensambles de la ACG de la Unidad 1 de la Central Nucleoelectrica Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Mijangos D, Z. E.; Vargas A, A. F.; Amador C, C., E-mail: zoedelfin@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    This work presents the determination of assemblies of the spent fuel pool (SFP) of the nuclear power plant of Laguna Verde (NPP-L V) which are candidates to be assigned to storage containers of independent spent fuel, with the objective of liberating decay heat and to have more space in the SFP, for the store of retired assemblies of the reactors in future reloads of NPP-L V, besides that the removed assemblies of the SFP should be stored in specific containers to guarantee the physical safety of them, as well as the radiological protection to the population and the environment. The design of the containers considered in this work is to store a maximum of 69 assemblies; it has a thermal capacity of 26 kilowatts and allows storing assemblies with a minimum of 5 years of have been extracted of the reactor core. Is considered that in 2016 start the storage of the spent assemblies on the containers, the candidates assemblies to store cover from the first reload in 1991, until the assemblies deposited in the SFP in the 14 reload in 2010; therefore in 2016, such assemblies will have fulfilled with the criteria of 5 years of have been removed of the Reactor, also the 69 assemblies assigned to each container will have a resulting decay heat that does not exceed the thermal capacity of the container, but that in great percentage approximates to the same one, and this way to take full advantage of their storage capacity and thermal capacity for each container. This work also contains the arrangement to accommodate the assemblies in the containers; such arrangement is constituted by areas according to the decay heat of each assembly. (Author)

  12. MELOX fuel fabrication plant: Operational feedback and future prospects

    International Nuclear Information System (INIS)

    Hugelmann, D.; Greneche, D.

    2000-01-01

    As of December 1, 1998, 32 Europeans LWRs are loaded with MOX fuel. It clearly means that plutonium recycling in MOX fuels is a mature industry, with successful operational experience in fabrication plants in some European countries, especially in France. Indeed, the recycling of plutonium generated in LWRs is one of the objectives of the full Reprocessing-Conditioning-Recycling (RCR) strategy chosen by France in the 70's. The most impressive results of this strategy, is the fact that 31 of the 32 reactors are loaded with MOX fuels supplied by the COGEMA Group from the same efficient fabrication process, the MIMAS process, improved for the MELOX plant to become the A-MIMAS process. In France, 17 reactors are already loaded and 11 additional reactors are technically suited to do so. Indeed, the EDF MOX program plans to use MOX in 28 of its 57 reactors. An EDF 900 MWe reactor core contains 157 assemblies of 264 rods each. 52 fuel assemblies per year are necessary for a 'UO 2 3-batches-MOX 3-batches' core management. In this case, a third of the UO 2 and a third of the MOX assemblies are replaced yearly, that means 36 UO 2 fuel assemblies and 16 MOX fuel assemblies. Some MOX fuelled reactors have now switched from the previously described core management to a so-called 'hybrid core management'. In this case, a quarter of UO 2 assemblies is replaced yearly. The first EDF reactor loaded with MOX fuel was Saint-Laurent B1, in 1987. The in-core experience, based on several hundred assemblies loaded, with reloading on a 1/3 cycle basis, shows that there is no operational difference between UO 2 and MOX fuels, both in terms of performance and safety. MOX fueling of 900 MWe EDF's PWRs, with a limited in-core MOX ratio of 30%, has needed only minor adaptations, such as addition of control rods, modification of the boron concentration in the cooling system and precaution against radiation exposure, easy to set up (optimisation of the fresh MOX fuel handling process, remote

  13. A unified methodology for single- and multiobjective in-core fuel management optimisation based on augmented Chebyshev scalarisation and a harmony search algorithm

    International Nuclear Information System (INIS)

    Schlünz, E.B.; Bokov, P.M.; Prinsloo, R.H.; Vuuren, J.H. van

    2016-01-01

    Highlights: • Unified methodology for in-core fuel management optimisation (ICFMO). • Addresses single- and multiobjective constrained and unconstrained ICFMO problems. • Augmented Chebyshev scalarising objective function with additive penalty function. • Harmony search algorithm yields high-quality solution or approximate Pareto set. • Methodology provides cycle-to-cycle optimisation decision support capabilities. - Abstract: The in-core fuel management optimisation (ICFMO) problem is the problem of finding an optimal fuel reload configuration for a nuclear reactor core. ICFMO may involve the pursuit of a single or multiple objectives, while satisfying several constraints. Very little multiobjective ICFMO research involving the fundamental notion of Pareto optimality has, however, been performed. In this paper, a unified methodology is proposed for the modelling and solution of single- and multiobjective ICFMO problems, be they constrained or unconstrained. With this methodology, ICFMO problems incorporating a variety of objectives and/or constraints may be modelled and solved rapidly, thus providing a cycle-to-cycle optimisation decision support capability for nuclear reactors. An augmented Chebyshev scalarising objective function is incorporated in the methodology for modelling any number of objectives, while an additive penalty function handles potential constraints. Furthermore, an adapted harmony search algorithm is used to solve a given ICFMO problem. The algorithm is able to yield a single solution or a nondominated set of solutions as result (depending on the number of objectives in a problem). The applicability of the methodology is demonstrated by solving (approximately) a variety of ICFMO test problems for the SAFARI-1 nuclear research reactor. The results indicate that the methodology may be used as an effective decision support tool for reactor operators tasked with designing reload configurations from cycle to cycle.

  14. Fuel trading

    International Nuclear Information System (INIS)

    2015-01-01

    A first part of this report proposes an overview of trends and predictions. After a synthesis on the sector changes and trends, it indicates and comments the most recent predictions for the consumption of refined oil products and for the turnover of the fuel wholesale market, reports the main highlights concerning the sector's life, and gives a dashboard of the sector activity. The second part proposes the annual report on trends and competition. It presents the main operator profiles and fuel categories, the main determining factors of the activity, the evolution of the sector context between 2005 and 2015 (consumptions, prices, temperature evolution). It analyses the evolution of the sector activity and indicators (sales, turnovers, prices, imports). Financial performances of enterprises are presented. The economic structure of the sector is described (evolution of the economic fabric, structural characteristics, French foreign trade). Actors are then presented and ranked in terms of turnover, of added value, and of result

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto.

    1991-01-01

    In a fuel assembly in which spectral shift type moderator guide members are arranged, the moderator guide member has a flow channel resistance member, that provides flow resistance against the moderators, in the upstream of a moderator flowing channel, by which the ratio of removing coolants is set greater at the upstream than downstream. With such a constitution, the void distribution increasing upward in the channel box except for the portion of the moderator guide member is moderated by the increase of the area of the void region that expands downward in the guide member. Accordingly, the axial power distribution is flattened throughout the operation cycle and excess distortion is eliminated to improve the fuel integrity. (T.M.)

  16. Fuel element

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1982-01-01

    Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  18. Fuel assemblies

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1983-01-01

    Purpose: To improve the operation performance of a BWR type reactor by improving the distribution of the uranium enrichment and the incorporation amount of burnable poisons in fuel assemblies. Constitution: The average enrichment of uranium 235 is increased in the upper portion as compared with that in the lower portion, while the incorporation amount of burnable poisons is increased in an upper portion as compared with that in the lower portion. The difference in the incorporation amount of the burnable poisons between the upper and lower portions is attained by charging two kinds of fuel rods; the ones incorporated with the burnable poisons over the entire length and the others incorporated with the burnable poisons only in the upper portions. (Seki, T.)

  19. Fuel assembly

    International Nuclear Information System (INIS)

    Hirukawa, Koji; Sakurada, Koichi.

    1992-01-01

    In a fuel assembly for a BWR type reactor, water rods or water crosses are disposed between fuel rods, and a value with a spring is disposed at the top of the coolant flow channel thereof, which opens a discharge port when pressure is increased to greater than a predetermined value. Further, a control element for the amount of coolant flow rate is inserted retractable to a control element guide tube formed at the lower portion of the water rod or the water cross. When the amount of control elements inserted to the control element guide tube is small and the inflown coolant flow rate is great, the void coefficient at the inside of the water rod is less than 5%. On the other hand, when the control elements are inserted, the flow resistance is increased, so that the void coefficient in the water rod is greater than 80%. When the pressure in the water rod is increased, the valve with the spring is raised to escape water or steams. Then, since the variation range of the change of the void coefficient can be controlled reliably by the amount of the control elements inserted, and nuclear fuel materials can be utilized effectively. (N.H.)

  20. Nonparametric combinatorial sequence models.

    Science.gov (United States)

    Wauthier, Fabian L; Jordan, Michael I; Jojic, Nebojsa

    2011-11-01

    This work considers biological sequences that exhibit combinatorial structures in their composition: groups of positions of the aligned sequences are "linked" and covary as one unit across sequences. If multiple such groups exist, complex interactions can emerge between them. Sequences of this kind arise frequently in biology but methodologies for analyzing them are still being developed. This article presents a nonparametric prior on sequences which allows combinatorial structures to emerge and which induces a posterior distribution over factorized sequence representations. We carry out experiments on three biological sequence families which indicate that combinatorial structures are indeed present and that combinatorial sequence models can more succinctly describe them than simpler mixture models. We conclude with an application to MHC binding prediction which highlights the utility of the posterior distribution over sequence representations induced by the prior. By integrating out the posterior, our method compares favorably to leading binding predictors.