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Sample records for brainpower fuels computer

  1. Even More Brain-Powered Science: Teaching and Learning with Discrepant Events. Brain-Powered Science Series

    Science.gov (United States)

    O'Brien, Thomas

    2011-01-01

    How can water and a penny demonstrate the power of mathematics and molecular theory? Do spelling and punctuation really matter to the human brain? The third of Thomas O'Brien's books designed for 5-12 grade science teachers, "Even More Brain-Powered Science" uses the questions above and 11 other inquiry-oriented discrepant events--experiments or…

  2. Ubiquitous human computing.

    Science.gov (United States)

    Zittrain, Jonathan

    2008-10-28

    Ubiquitous computing means network connectivity everywhere, linking devices and systems as small as a drawing pin and as large as a worldwide product distribution chain. What could happen when people are so readily networked? This paper explores issues arising from two possible emerging models of ubiquitous human computing: fungible networked brainpower and collective personal vital sign monitoring.

  3. Fuel element performance computer modelling

    International Nuclear Information System (INIS)

    Locke, D.H.

    1978-01-01

    The meeting was attended by 88 participants from 17 countries. Altogether 47 papers were presented. The majority of the presentations contained a description of the equations and solutions used to describe and evaluate some of the physical processes taking place in water reactor fuel pins under irradiation. At the same time, particular attention was paid to the ''bench marking'' of the codes wherein solutions arrived at for particular experiments are compared with the results at the experiments

  4. The Relieving Effects of BrainPower Advanced, a Dietary Supplement, in Older Adults with Subjective Memory Complaints: A Randomized, Double-Blind, Placebo-Controlled Trial

    Directory of Open Access Journals (Sweden)

    Jingfen Zhu

    2016-01-01

    Full Text Available Subjective memory complaints (SMCs are common in older adults that can often predict further cognitive impairment. No proven effective agents are available for SMCs. The effect of BrainPower Advanced, a dietary supplement consisting of herbal extracts, nutrients, and vitamins, was evaluated in 98 volunteers with SMCs, averaging 67 years of age (47–88, in a randomized, double-blind, placebo-controlled trial. Subjective hypomnesis/memory loss (SML and attention/concentration deficits (SAD were evaluated before and after 12-week supplementation of BrainPower Advanced capsules (n=47 or placebo (n=51, using a 5-point memory questionnaire (1 = no/slight, 5 = severe. Objective memory function was evaluated using 3 subtests of visual/audio memory, abstraction, and memory recall that gave a combined total score. The BrainPower Advanced group had more cases of severe SML (severity ⩾ 3 (44/47 and severe SAD (43/47 than the placebo group (39/51 and 37/51, < 0.05, < 0.05, resp. before the treatment. BrainPower Advanced intervention, however, improved a greater proportion of the severe SML (29.5%(13/44 (P<0.01 and SAD (34.9%(15/43(P<0.01 than placebo (5.1% (2/39 and 13.5% (5/37, resp.. Thus, 3-month BrainPower Advanced supplementation appears to be beneficial to older adults with SMCs.

  5. The Relieving Effects of BrainPower Advanced, a Dietary Supplement, in Older Adults with Subjective Memory Complaints: A Randomized, Double-Blind, Placebo-Controlled Trial.

    Science.gov (United States)

    Zhu, Jingfen; Shi, Rong; Chen, Su; Dai, Lihua; Shen, Tian; Feng, Yi; Gu, Pingping; Shariff, Mina; Nguyen, Tuong; Ye, Yeats; Rao, Jianyu; Xing, Guoqiang

    2016-01-01

    Subjective memory complaints (SMCs) are common in older adults that can often predict further cognitive impairment. No proven effective agents are available for SMCs. The effect of BrainPower Advanced, a dietary supplement consisting of herbal extracts, nutrients, and vitamins, was evaluated in 98 volunteers with SMCs, averaging 67 years of age (47-88), in a randomized, double-blind, placebo-controlled trial. Subjective hypomnesis/memory loss (SML) and attention/concentration deficits (SAD) were evaluated before and after 12-week supplementation of BrainPower Advanced capsules (n = 47) or placebo (n = 51), using a 5-point memory questionnaire (1 = no/slight, 5 = severe). Objective memory function was evaluated using 3 subtests of visual/audio memory, abstraction, and memory recall that gave a combined total score. The BrainPower Advanced group had more cases of severe SML (severity ⩾ 3) (44/47) and severe SAD (43/47) than the placebo group (39/51 and 37/51, < 0.05, < 0.05, resp.) before the treatment. BrainPower Advanced intervention, however, improved a greater proportion of the severe SML (29.5%)(13/44) (P < 0.01) and SAD (34.9%)(15/43)(P < 0.01) than placebo (5.1% (2/39) and 13.5% (5/37), resp.). Thus, 3-month BrainPower Advanced supplementation appears to be beneficial to older adults with SMCs.

  6. Computational Design of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-01-01

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  7. Resonance computations for cells with fuel annuli

    International Nuclear Information System (INIS)

    Hwang, R.N.; Gelbard, E.M.

    1990-01-01

    Two methods have been developed for the computation of resonance integrals in cells containing annular fuel regions. Both are based on rational approximations. One is a generalization of a one-term rational approximation method developed by Segev for a cell with a single fuel annulus. The second modifies the earlier Chen-Gelbard two-term method originally used for double-heterogeneity calculations. Both methods were tested, in cells with two fuel annuli, for various U 235 and U 238 resonances. Both gives resonance integrals accurate enough for practical purposes. The two-term fits are substantially more accurate in some NR cases, but are somewhat more difficult to correct for finite resonance widths. 8 refs., 4 tabs

  8. How to harvest brainpower for a brighter future on Earth - bypassing stereotypes in practice

    Science.gov (United States)

    Hammarlund, Emma U.

    2016-04-01

    Our ability to detect and respond to patterns, the core asset in scientific work, likely has been a strong evolutionary advantage when navigating between need and danger. We therefore argue that our use of stereotypes for the most part is decoupled from malice but instead profoundly rooted, pervasive and often unchallenged in our daily thought process. The fundamental need for patterns also means that we all express biases - regardless of our gender - such as during teaching, mentoring or collaborations. As culprits, our implicit bias both disfavor (e.g. female or immigrant students) and favor (e.g. the-celebrated-worker's-results-are-probably-stronger-than-they-seem) our fellow workers in unsolicited ways. Additionally, people in our environment adapt to the expectations we express, called the stereotype threat. Ultimately, these two effects likely lead to missed opportunities for new questions and achievements. The omni-present implicit bias also means that some equality efforts, such as quotas, would soon be neutralized unless we train in catching our use of stereotypes in action. We will share our experiences from gender sensitivity training during teaching and assessment, such as the use of anonymous exams, simple 'traps' to demonstrate gender bias, and group work management. We also present views collected in our group on how academic talent management may improve through mentorship, career coaching and general policy decisions. All with the selfish ambition to improve future harvests of brainpower, for science and for sustainable human life on Earth.

  9. Computer-assisted nuclear fuel manufacture

    International Nuclear Information System (INIS)

    Maloney, J.P.; Schaumann, S.M.; Stone, E.

    1976-01-01

    At the ERDA Savannah River Plant, a process monitor, which incorporates an online digital computer, assists in manufacturing fuel elements used to produce nuclides such as plutonium, tritium, and californium in the plant's nuclear reactors. Also, inventory functions assist in safeguarding fissile material and protecting against accidental nuclear criticality. Terminals at strategic locations throughout the process area enable production operators to send and receive instructions and information on each manufacturing step

  10. Computer-assisted nuclear fuel manufacture

    International Nuclear Information System (INIS)

    Maloney, J.P.; Schaumann, C.M.; Stone, E.

    1976-06-01

    At the ERDA Savannah River Plant, a process monitor, which incorporates an online digital computer, assists in manufacturing fuel elements used to produce nuclides such as plutonium, tritium, and californium in the plant's nuclear reactors. Also, inventory functions assist in safeguarding fissile material and protecting against accidental nuclear criticality. Terminals at strategic locations throughout the process area enable production operators to send and receive instructions and information on each manufacturing step. 11 fig

  11. Computer simulation of variform fuel assemblies using Dragon code

    International Nuclear Information System (INIS)

    Ju Haitao; Wu Hongchun; Yao Dong

    2005-01-01

    The DRAGON is a cell code that developed for the CANDU reactor by the Ecole Polytechnique de Montreal of CANADA. Although, the DRAGON is mainly used to simulate the CANDU super-cell fuel assembly, it has an ability to simulate other geometries of the fuel assembly. However, only NEACRP benchmark problem of the BWR lattice cell was analyzed until now except for the CANDU reactor. We also need to develop the code to simulate the variform fuel assemblies, especially, for design of the advanced reactor. We validated that the cell code DRAGON is useful for simulating various kinds of the fuel assembly by analyzing the rod-shape fuel assembly of the PWR and the MTR plate-shape fuel assembly. Some other kinds of geometry of geometry were computed. Computational results show that the DRAGON is able to analyze variform fuel assembly problems and the precision is high. (authors)

  12. MISER-I: a computer code for JOYO fuel management

    International Nuclear Information System (INIS)

    Yamashita, Yoshioki

    1976-06-01

    A computer code ''MISER-I'' is for a nuclear fuel management of Japan Experimental Fast Breeder Reactor JOYO. The nuclear fuel management in JOYO can be regarded as a fuel assembly management because a handling unit of fuel in JOYO plant is a fuel subassembly (core and blanket subassembly), and so the recording of material balance in computer code is made with each subassembly. The input information into computer code is given with each subassembly for a transfer operation, or with one reactor cycle and every one month for a burn-up in reactor core. The output information of MISER-I code is the fuel assembly storage record, fuel storage weight record in each material balance subarea at any specified day, and fuel subassembly transfer history record. Change of nuclear fuel composition and weight due to a burn-up is calculated with JOYO-Monitoring Code by off-line computation system. MISER-I code is written in FORTRAN-IV language for FACOM 230-48 computer. (auth.)

  13. The Relieving Effects of BrainPower Advanced, a Dietary Supplement, in Older Adults with Subjective Memory Complaints: A Randomized, Double-Blind, Placebo-Controlled Trial

    OpenAIRE

    Zhu, Jingfen; Shi, Rong; Chen, Su; Dai, Lihua; Shen, Tian; Feng, Yi; Gu, Pingping; Shariff, Mina; Nguyen, Tuong; Ye, Yeats; Rao, Jianyu; Xing, Guoqiang

    2016-01-01

    Subjective memory complaints (SMCs) are common in older adults that can often predict further cognitive impairment. No proven effective agents are available for SMCs. The effect of BrainPower Advanced, a dietary supplement consisting of herbal extracts, nutrients, and vitamins, was evaluated in 98 volunteers with SMCs, averaging 67 years of age (47?88), in a randomized, double-blind, placebo-controlled trial. Subjective hypomnesis/memory loss (SML) and attention/concentration deficits (SAD) w...

  14. Computer simulation of fuel element performance

    Energy Technology Data Exchange (ETDEWEB)

    Sukhanov, G I

    1979-01-01

    The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

  15. Spent fuel management fee methodology and computer code user's manual

    International Nuclear Information System (INIS)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively

  16. Advanced technique for computing fuel combustion properties in pulverized-fuel fired boilers

    Energy Technology Data Exchange (ETDEWEB)

    Kotler, V.R. (Vsesoyuznyi Teplotekhnicheskii Institut (Russian Federation))

    1992-03-01

    Reviews foreign technical reports on advanced techniques for computing fuel combustion properties in pulverized-fuel fired boilers and analyzes a technique developed by Combustion Engineering, Inc. (USA). Characteristics of 25 fuel types, including 19 grades of coal, are listed along with a diagram of an installation with a drop tube furnace. Characteristics include burn-out intensity curves obtained using thermogravimetric analysis for high-volatile bituminous, semi-bituminous and coking coal. The patented LFP-SKM mathematical model is used to model combustion of a particular fuel under given conditions. The model allows for fuel particle size, air surplus, load, flame height, and portion of air supplied as tertiary blast. Good agreement between computational and experimental data was observed. The method is employed in designing new boilers as well as converting operating boilers to alternative types of fuel. 3 refs.

  17. Fuel rod computations. The COMETHE code in its CEA version

    International Nuclear Information System (INIS)

    Lenepveu, Dominique.

    1976-01-01

    The COMETHE code (COde d'evolution MEcanique et THermique) is intended for computing the irradiation behavior of water reactor fuel pins. It is concerned with steadily operated cylindrical pins, containing fuel pellet stacks (UO 2 or PuO 2 ). The pin consists in five different axial zones: two expansion chambers, two blankets, and a central core that may be divided into several stacks parted by plugs. As far as computation is concerned, the pin is divided into slices (maximum 15) in turn divided into rings (maximum 50). Information are obtained for each slice: the radial temperature distribution, heat transfer coefficients, thermal flux at the pin surface, changes in geometry according to temperature conditions, and specific burn-up. The physical models involved take account for: heat transfer, fission gas release, fuel expansion, and creep of the can. Results computed with COMETHE are compared with those from ELP and EPEL irradiation experiments [fr

  18. Neutron physics computation of CERCA fuel elements for Maria Reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

    2008-01-01

    Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

  19. Emission computer tomography on a Dodewaard mixed oxide fuel pin

    International Nuclear Information System (INIS)

    Buurveld, H.A.; Dassel, G.

    1993-12-01

    A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

  20. Computational methods for more fuel-efficient ship

    NARCIS (Netherlands)

    Koren, B.

    2008-01-01

    The flow of water around a ship powered by a combustion engine is a key factor in the ship's fuel consumption. The simulation of flow patterns around ship hulls is therefore an important aspect of ship design. While lengthy computations are required for such simulations, research by Jeroen Wackers

  1. 35-We polymer electrolyte membrane fuel cell system for notebook computer using a compact fuel processor

    Science.gov (United States)

    Son, In-Hyuk; Shin, Woo-Cheol; Lee, Yong-Kul; Lee, Sung-Chul; Ahn, Jin-Gu; Han, Sang-Il; kweon, Ho-Jin; Kim, Ju-Yong; Kim, Moon-Chan; Park, Jun-Yong

    A polymer electrolyte membrane fuel cell (PEMFC) system is developed to power a notebook computer. The system consists of a compact methanol-reforming system with a CO preferential oxidation unit, a 16-cell PEMFC stack, and a control unit for the management of the system with a d.c.-d.c. converter. The compact fuel-processor system (260 cm 3) generates about 1.2 L min -1 of reformate, which corresponds to 35 We, with a low CO concentration (notebook computers.

  2. 35-We polymer electrolyte membrane fuel cell system for notebook computer using a compact fuel processor

    Energy Technology Data Exchange (ETDEWEB)

    Son, In-Hyuk; Shin, Woo-Cheol; Lee, Sung-Chul; Ahn, Jin-Gu; Han, Sang-Il; kweon, Ho-Jin; Kim, Ju-Yong; Park, Jun-Yong [Energy 1 Group, Energy Laboratory at Corporate R and D Center in Samsung SDI Co., Ltd., 575, Shin-dong, Yeongtong-gu, Suwon-si, Gyeonggi-do 443-731 (Korea); Lee, Yong-Kul [Department of Chemical Engineering, Dankook University, Youngin 448-701 (Korea); Kim, Moon-Chan [Department of Environmental Engineering, Chongju University, Chongju 360-764 (Korea)

    2008-10-15

    A polymer electrolyte membrane fuel cell (PEMFC) system is developed to power a notebook computer. The system consists of a compact methanol-reforming system with a CO preferential oxidation unit, a 16-cell PEMFC stack, and a control unit for the management of the system with a d.c.-d.c. converter. The compact fuel-processor system (260 cm{sup 3}) generates about 1.2 L min{sup -1} of reformate, which corresponds to 35 We, with a low CO concentration (<30 ppm, typically 0 ppm), and is thus proven to be capable of being targetted at notebook computers. (author)

  3. Evolution of the Darlington NGS fuel handling computer systems

    International Nuclear Information System (INIS)

    Leung, V.; Crouse, B.

    1996-01-01

    The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs

  4. Evolution of the Darlington NGS fuel handling computer systems

    Energy Technology Data Exchange (ETDEWEB)

    Leung, V; Crouse, B [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station

    1997-12-31

    The ability to improve the capabilities and reliability of digital control systems in nuclear power stations to meet changing plant and personnel requirements is a formidable challenge. Many of these systems have high quality assurance standards that must be met to ensure adequate nuclear safety. Also many of these systems contain obsolete hardware along with software that is not easily transported to newer technology computer equipment. Combining modern technology upgrades into a system of obsolete hardware components is not an easy task. Lastly, as users become more accustomed to using modern technology computer systems in other areas of the station (e.g. information systems), their expectations of the capabilities of the plant systems increase. This paper will present three areas of the Darlington NGS fuel handling computer system that have been or are in the process of being upgraded to current technology components within the framework of an existing fuel handling control system. (author). 3 figs.

  5. CRACKEL: a computer code for CFR fuel management calculations

    International Nuclear Information System (INIS)

    Burstall, R.F.; Ball, M.A.; Thornton, D.E.J.

    1975-12-01

    The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)

  6. Computer imaging of EBR-II fuel handling equipment

    International Nuclear Information System (INIS)

    Peters, G.G.; Hansen, L.H.

    1995-01-01

    This paper describes a three-dimensional graphics application used to visualize the positions of remotely operated fuel handling equipment in the EBR-II reactor. A three-dimensional (3D) visualization technique is necessary to simulate direct visual observation of the transfers of fuel and experiments into and out of the reactor because the fuel handling equipment is submerged in liquid sodium and therefore is not visible to the operator. The system described in this paper uses actual signals to drive a three-dimensional computer-generated model in real-time in response to movements of equipment in the plant This paper will present details on how the 3D model of the intank equipment was created and how real-time dynamic behavior was added to each of the moving components

  7. COMTA - a computer code for fuel mechanical and thermal analysis

    International Nuclear Information System (INIS)

    Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.

    1979-01-01

    COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)

  8. Workstation computer systems for in-core fuel management

    International Nuclear Information System (INIS)

    Ciccone, L.; Casadei, A.L.

    1992-01-01

    The advancement of powerful engineering workstations has made it possible to have thermal-hydraulics and accident analysis computer programs operating efficiently with a significant performance/cost ratio compared to large mainframe computer. Today, nuclear utilities are acquiring independent engineering analysis capability for fuel management and safety analyses. Computer systems currently available to utility organizations vary widely thus requiring that this software be operational on a number of computer platforms. Recognizing these trends Westinghouse adopted a software development life cycle process for the software development activities which strictly controls the development, testing and qualification of design computer codes. In addition, software standards to ensure maximum portability were developed and implemented, including adherence to FORTRAN 77, and use of uniform system interface and auxiliary routines. A comprehensive test matrix was developed for each computer program to ensure that evolution of code versions preserves the licensing basis. In addition, the results of such test matrices establish the Quality Assurance basis and consistency for the same software operating on different computer platforms. (author). 4 figs

  9. Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, James H. [University of North Florida; Cox, Philip [University of North Florida; Harrington, William J [University of North Florida; Campbell, Joseph L [University of North Florida

    2013-09-03

    ABSTRACT Project Title: Recovery Act: Advanced Direct Methanol Fuel Cell for Mobile Computing PROJECT OBJECTIVE The objective of the project was to advance portable fuel cell system technology towards the commercial targets of power density, energy density and lifetime. These targets were laid out in the DOE’s R&D roadmap to develop an advanced direct methanol fuel cell power supply that meets commercial entry requirements. Such a power supply will enable mobile computers to operate non-stop, unplugged from the wall power outlet, by using the high energy density of methanol fuel contained in a replaceable fuel cartridge. Specifically this project focused on balance-of-plant component integration and miniaturization, as well as extensive component, subassembly and integrated system durability and validation testing. This design has resulted in a pre-production power supply design and a prototype that meet the rigorous demands of consumer electronic applications. PROJECT TASKS The proposed work plan was designed to meet the project objectives, which corresponded directly with the objectives outlined in the Funding Opportunity Announcement: To engineer the fuel cell balance-of-plant and packaging to meet the needs of consumer electronic systems, specifically at power levels required for mobile computing. UNF used existing balance-of-plant component technologies developed under its current US Army CERDEC project, as well as a previous DOE project completed by PolyFuel, to further refine them to both miniaturize and integrate their functionality to increase the system power density and energy density. Benefits of UNF’s novel passive water recycling MEA (membrane electrode assembly) and the simplified system architecture it enabled formed the foundation of the design approach. The package design was hardened to address orientation independence, shock, vibration, and environmental requirements. Fuel cartridge and fuel subsystems were improved to ensure effective fuel

  10. Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

    International Nuclear Information System (INIS)

    Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

    2002-01-01

    The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

  11. 10 CFR Appendix II to Part 504 - Fuel Price Computation

    Science.gov (United States)

    2010-01-01

    ... DEPARTMENT OF ENERGY (CONTINUED) ALTERNATE FUELS EXISTING POWERPLANTS Pt. 504, App. II Appendix II to Part... effects of future real price increases for each fuel. The delivered price of an alternate fuel used to calculate delivered fuel expenses must reflect the petitioner's delivered price of the alternate fuel and...

  12. Borrowing brainpower - sharing insecurities

    DEFF Research Database (Denmark)

    Wegener, Charlotte; Meier, Ninna; Ingerslev, Karen

    2016-01-01

    Academic writing is a vital, yet complex skill that must be developed within a doctoral training process. In addition, becoming an academic researcher is a journey of changing sense of self and identity. Through analysis of a group session, we show how the feedback of peers addresses questions...... of structure and writing style along with wider issues of researcher identity. Thus, peer learning is demonstrated as a process of simultaneously building a text and an identity as scholarly researcher. The paper advocates ‘borrowing brainpower’ from peers in order to write better texts and, at the same time......, ‘share insecurities’ during the development of the researcher identity. Based on a distributed notion of peer learning and identity, we point to the need for further research into the everyday activities of doctoral writing groups in order to understand the dynamic relationship between production of text...

  13. MEASUREMENTS AND COMPUTATIONS OF FUEL DROPLET TRANSPORT IN TURBULENT FLOWS

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Katz and Omar Knio

    2007-01-10

    The objective of this project is to study the dynamics of fuel droplets in turbulent water flows. The results are essential for development of models capable of predicting the dispersion of slightly light/heavy droplets in isotropic turbulence. Since we presently do not have any experimental data on turbulent diffusion of droplets, existing mixing models have no physical foundations. Such fundamental knowledge is essential for understanding/modeling the environmental problems associated with water-fuel mixing, and/or industrial processes involving mixing of immiscible fluids. The project has had experimental and numerical components: 1. The experimental part of the project has had two components. The first involves measurements of the lift and drag forces acting on a droplet being entrained by a vortex. The experiments and data analysis associated with this phase are still in progress, and the facility, constructed specifically for this project is described in Section 3. In the second and main part, measurements of fuel droplet dispersion rates have been performed in a special facility with controlled isotropic turbulence. As discussed in detail in Section 2, quantifying and modeling the of droplet dispersion rate requires measurements of their three dimensional trajectories in turbulent flows. To obtain the required data, we have introduced a new technique - high-speed, digital Holographic Particle Image Velocimetry (HPIV). The technique, experimental setup and results are presented in Section 2. Further information is available in Gopalan et al. (2005, 2006). 2. The objectives of the numerical part are: (1) to develop a computational code that combines DNS of isotropic turbulence with Lagrangian tracking of particles based on integration of a dynamical equation of motion that accounts for pressure, added mass, lift and drag forces, (2) to perform extensive computations of both buoyant (bubbles) and slightly buoyant (droplets) particles in turbulence conditions

  14. A computational methodology for formulating gasoline surrogate fuels with accurate physical and chemical kinetic properties

    KAUST Repository

    Ahmed, Ahfaz; Goteng, Gokop; Shankar, Vijai; Al-Qurashi, Khalid; Roberts, William L.; Sarathy, Mani

    2015-01-01

    simpler molecular composition that represent real fuel behavior in one or more aspects are needed to enable repeatable experimental and computational combustion investigations. This study presents a novel computational methodology for formulating

  15. Computer-controlled gamma-ray scanner for irradiated reactor fuel

    International Nuclear Information System (INIS)

    Mandler, J.W.; Coates, R.A.; Killian, E.W.

    1979-01-01

    Gamma-ray scanning of irradiated fuel is an important nondestructive technique used in the thermal fuels behavior program currently under way at the Idaho National Engineering Laboratory. This paper is concerned with the computer-controlled isotopic gamma-ray-scanning system developed for postirradiation examination of fuel and includes a brief discussion of some scan results obtained from fuel rods irradiated in the Power-Burst Facility to illustrate gamma-ray spectrometry for this application. Both burnup profiles and information concerning fission-product migration in irradiated fuel are routinely obtained with the computer-controlled system

  16. A New Methodology for Fuel Mass Computation of an operating Aircraft

    Directory of Open Access Journals (Sweden)

    M Souli

    2016-03-01

    Full Text Available The paper performs a new computational methodology for an accurate computation of fuel mass inside an aircraft wing during the flight. The computation is carried out using hydrodynamic equations, classically known as Navier-Stokes equations by the CFD community. For this purpose, a computational software is developed, the software computes the fuel mass inside the tank based on experimental data of pressure gages that are inserted in the fuel tank. Actually and for safety reasons, Optical fiber sensor for fluid level sensor detection is used. The optical system consists to an optically controlled acoustic transceiver system which measures the fuel level inside the each compartment of the fuel tank. The system computes fuel volume inside the tank and needs density to compute the total fuel mass. Using optical sensor technique, density measurement inside the tank is required. The method developed in the paper, requires pressure measurements in each tank compartment, the density is then computed based on pressure measurements and hydrostatic assumptions. The methodology is tested using a fuel tank provided by Airbus for time history refueling process.

  17. Performance Evaluation of Solid Oxide Fuel Cell by Computer ...

    African Journals Online (AJOL)

    The search for sustainable energy source that can compete with the existing one led to the discovery and acceptance of fuel cell technologies as a perfect replacement for fossil fuel. The ability of Solid Oxide Fuel Cells (SOFC) to capture the heat generation during the process of energy generation from electrochemical ...

  18. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  19. The failure mechanisms of HTR coated particle fuel and computer code

    International Nuclear Information System (INIS)

    Yang Lin; Liu Bing; Shao Youlin; Liang Tongxiang; Tang Chunhe

    2010-01-01

    The basic constituent unit of fuel element in HTR is ceramic coated particle fuel. And the performance of coated particle fuel determines the safety of HTR. In addition to the traditional detection of radiation experiments, establishing computer code is of great significance to the research. This paper mainly introduces the structure and the failure mechanism of TRISO-coated particle fuel, as well as a few basic assumptions,principles and characteristics of some existed main overseas codes. Meanwhile, this paper has proposed direction of future research by comparing the advantages and disadvantages of several computer codes. (authors)

  20. Fuel behavior modeling using the MARS computer code

    International Nuclear Information System (INIS)

    Faya, S.C.S.; Faya, A.J.G.

    1983-01-01

    The fuel behaviour modeling code MARS against experimental data, was evaluated. Two cases were selected: an early comercial PWR rod (Maine Yankee rod) and an experimental rod from the Canadian BWR program (Canadian rod). The MARS predictions are compared with experimental data and predictions made by other fuel modeling codes. Improvements are suggested for some fuel behaviour models. Mars results are satisfactory based on the data available. (Author) [pt

  1. 3-Dimensional Computational Fluid Dynamics Modeling of Solid Oxide Fuel Cell Using Different Fuels

    Science.gov (United States)

    2011-01-01

    major types of fuel cells in practice are listed below: Polymer Electrolyte Membrane Fuel Cell ( PEMFC ) Alkaline Fuel cell (AFC) Phosphoric Acid...Material Operating Temperature (oC) Efficiency (%) PEMFC H2, Methanol, Formic Acid Hydrated Organic Polymer < 90 40-50 AFC Pure H2 Aqueous

  2. PAPIRUS - a computer code for FBR fuel performance analysis

    International Nuclear Information System (INIS)

    Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

    1991-01-01

    The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

  3. Design and construction of wireless personal computer based fuel ...

    African Journals Online (AJOL)

    Results of the different fuel levels obtained through wireless medium are graphical represented on the PC display for monitoring and prompt actions. The PC interface was made possible using C++ programming language. Keywords: Monitoring system, fuel level, floater, transceiver and Radio frequency. Nigerian Journal of ...

  4. A computational methodology for formulating gasoline surrogate fuels with accurate physical and chemical kinetic properties

    KAUST Repository

    Ahmed, Ahfaz

    2015-03-01

    Gasoline is the most widely used fuel for light duty automobile transportation, but its molecular complexity makes it intractable to experimentally and computationally study the fundamental combustion properties. Therefore, surrogate fuels with a simpler molecular composition that represent real fuel behavior in one or more aspects are needed to enable repeatable experimental and computational combustion investigations. This study presents a novel computational methodology for formulating surrogates for FACE (fuels for advanced combustion engines) gasolines A and C by combining regression modeling with physical and chemical kinetics simulations. The computational methodology integrates simulation tools executed across different software platforms. Initially, the palette of surrogate species and carbon types for the target fuels were determined from a detailed hydrocarbon analysis (DHA). A regression algorithm implemented in MATLAB was linked to REFPROP for simulation of distillation curves and calculation of physical properties of surrogate compositions. The MATLAB code generates a surrogate composition at each iteration, which is then used to automatically generate CHEMKIN input files that are submitted to homogeneous batch reactor simulations for prediction of research octane number (RON). The regression algorithm determines the optimal surrogate composition to match the fuel properties of FACE A and C gasoline, specifically hydrogen/carbon (H/C) ratio, density, distillation characteristics, carbon types, and RON. The optimal surrogate fuel compositions obtained using the present computational approach was compared to the real fuel properties, as well as with surrogate compositions available in the literature. Experiments were conducted within a Cooperative Fuels Research (CFR) engine operating under controlled autoignition (CAI) mode to compare the formulated surrogates against the real fuels. Carbon monoxide measurements indicated that the proposed surrogates

  5. The computer program ELCOM in the planning and structural analysis of PWR fuel elements: an example

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da

    1990-01-01

    Is's presented some results obtained with the ELCOM computer code, such as deflections, moments and natural frequencies, used in the design and structural analysis of PWR fuels assemblies. It's studied the behavior of these results varying the number of spacer grids, the rigidity of the joint between the fuel pin and the spacer grid, and the fuel assembly's boundary condition, considered in the analysis, in it's mounting into the core (if clamped-clamped, clamped-hinged or hinged-hinged). (author)

  6. Computer simulation of HTGR fuel microspheres using a Monte-Carlo statistical approach

    International Nuclear Information System (INIS)

    Hedrick, C.E.

    1976-01-01

    The concept and computational aspects of a Monte-Carlo statistical approach in relating structure of HTGR fuel microspheres to the uranium content of fuel samples have been verified. Results of the preliminary validation tests and the benefits to be derived from the program are summarized

  7. KWIKPLAN: a computer program for projecting the annual requirements of nuclear fuel cycle operations

    International Nuclear Information System (INIS)

    Salmon, R.; Kee, C.W.

    1977-06-01

    The computer code KWIKPLAN was written to facilitate the calculation of projected nuclear fuel cycle activities. Using given projections of power generation, the code calculates annual requirements for fuel fabrication, fuel reprocessing, uranium mining, and plutonium use and production. The code uses installed capacity projections and mass flow data for six types of reactors to calculate projected fuel cycle activities and inventories. It calculates fissile uranium and plutonium flows and inventories after allowing for an economy with limited reprocessing capacity and a backlog of unreprocessed fuel. All calculations are made on a quarterly basis; printed and punched output of the projected fuel cycle activities are made on an annual basis. Since the punched information is used in another code to determine waste inventories, the code punches a table from which the effective average burnup can be calculated for the fuel being reprocessed

  8. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  9. A computer program for structural analysis of fuel elements

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-01-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equivalent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

  10. Development of a computer program for the cost analysis of spent fuel management

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won; Cha, Jeong Hun; Whang, Joo Ho

    2009-01-01

    So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

  11. TEMP: a computer code to calculate fuel pin temperatures during a transient

    International Nuclear Information System (INIS)

    Bard, F.E.; Christensen, B.Y.; Gneiting, B.C.

    1980-04-01

    The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method

  12. Calculation of demands for nuclear fuels and fuel cycle services. Description of computer model and strategies developed by Working Group 1

    International Nuclear Information System (INIS)

    Working Group 1 examined a range of reactor deployment strategies and fuel cycle options, in oder to estimate the range of nuclear fuel requirements and fuel cycle service needs which would result. The computer model, its verification in comparison with other models, the strategies to be examined through use of the model, and the range of results obtained are described

  13. Development of a Computer Code for the Estimation of Fuel Rod Failure

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, I.H.; Ahn, H.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    Much research has already been performed to obtain the information on the degree of failed fuel rods from the primary coolant activities of operating PWRs in the last few decades. The computer codes that are currently in use for domestic nuclear power plants, such as CADE code and ABB-CE codes developed by Westinghouse and ABB-CE, respectively, still give significant overall errors in estimating the failed fuel rods. In addition, with the CADE code, it is difficult to predict the degree of fuel rod failures during the transient period of nuclear reactor operation, where as the ABB-CE codes are relatively more difficult to use for end-users. In particular, the rapid progresses made recently in the area of the computer hardware and software systems that their computer programs be more versatile and user-friendly. While the MS windows system that is centered on the graphic user interface and multitasking is now in widespread use, the computer codes currently employed at the nuclear power plants, such as CADE and ABB-CE codes, can only be run on the DOS system. Moreover, it is desirable to have a computer code for the fuel rod failure estimation that can directly use the radioactivity data obtained from the on-line monitoring system of the primary coolant activity. The main purpose of this study is, therefore, to develop a Windows computer code that can predict the location, the number of failed fuel rods,and the degree of failures using the radioactivity data obtained from the primary coolant activity for PWRs. Another objective is to combine this computer code with the on-line monitoring system of the primary coolant radioactivity at Kori 3 and 4 operating nuclear power plants and enable their combined use for on-line evaluation of the number and degree of fuel rod failures. (author). 49 refs., 85 figs., 30 tabs.

  14. COMETHE III J a computer code for predicting mechanical and thermal behaviour of a fuel pin

    International Nuclear Information System (INIS)

    Verbeek, P.; Hoppe, N.

    1976-01-01

    The design of fuel pins for power reactors requires a realistic evaluation of their thermal and mechanical performances throughout their irradiation life. This evaluation involves the knowledge of a number of parameters, very intricate and interconnected, for example, the temperature, the restructuring and the swelling rates of the fuel pellets, the dimensions, the stresses and the strains in the clad, the composition and the properties of gases, the inner gas pressure etc. This complex problem can only be properly handled by a computer programme which analyses the fuel pin thermal and mechanical behaviour at successive steps of its irradiation life. This report presents an overall description of the COMETHE III-J computer programme, designed to calculate the integral performance of oxide fuel pins with cylindrical metallic cladding irradiated in thermal or fast flux. (author)

  15. Specialists' meeting on fuel element performance computer modelling, Preston, United Kingdom, 15-19 March 1982

    International Nuclear Information System (INIS)

    1983-03-01

    The 46 papers of the meeting concerned with computer models of Water Reactor fuel elements cover practically all aspects of behavior of fuel elements in normal operation and in accident condition. Each session of the meeting produced a critical evaluation of one of the 5 topics into which the subject area had been divided. The sessions' report summarize the papers and make recommendations for further work. Separate abstracts were prepared for all the papers presented at this meeting

  16. Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

    International Nuclear Information System (INIS)

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

    1976-01-01

    Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) ' [es

  17. Development, verification and validation of the fuel channel behaviour computer code FACTAR

    Energy Technology Data Exchange (ETDEWEB)

    Westbye, C J; Brito, A C; MacKinnon, J C; Sills, H E; Langman, V J [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thermal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate loss of coolant accident conditions including transition and large break LOCA`s (loss of coolant accidents) with emergency coolant injection assumed available. FACTAR`s predictions of fuel temperature and sheath failure times are used to subsequent assessment of fission product releases and fuel string expansion. This paper discusses the origin and development history of FACTAR, presents the mathematical models and solution technique, the detailed quality assurance procedures that are followed during development, and reports the future development of the code. (author). 27 refs., 3 figs.

  18. Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

    International Nuclear Information System (INIS)

    Khalil, H.

    1988-01-01

    An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

  19. PWR fuel physico chemistry. Improvements of the Sage code to compute thermochemical balance in PWR fuel

    International Nuclear Information System (INIS)

    Garcia, P.; Baron, D.; Piron, J.P.

    1993-02-01

    A physicochemical survey of high burnup fuel has been undertaken in the context of a 3-party action (CEA Cadarache - EDF/DER - FRAMATOME). One of the tasks involved consists in adapting the SAGE code for assessment of the thermochemical equilibria of fission products in solution in the fuel matrix. This paper describes the first stage of this task. Even if other improvements are planned, the oxid oxygen potentials are yet properly reproduced for the simulated burnup. (authors). 63 figs., 4 tabs., 41 refs

  20. Computed tomography on a defective CANDU fuel pencil end cap

    International Nuclear Information System (INIS)

    Lupton, L.R.

    1985-09-01

    Five tomographic slices through a defective end cap from a CANDU fuel pencil have been generated using a Co-60 source and a first generation translate-rotate tomography scanner. An anomaly in the density distribution that is believed to have resulted from the defect has been observed. However, with the 0.30 mm spatial resolution used, it has not been possible to state unequivocally whether the change in density is caused by a defect in the weld or a statistical anomaly in the data. It is concluded that a microtomography system, with a spatial resolution in the range of 0.1 mm, could detect the flaw

  1. The GC computer code for flow sheet simulation of pyrochemical processing of spent nuclear fuels

    International Nuclear Information System (INIS)

    Ahluwalia, R.K.; Geyer, H.K.

    1996-01-01

    The GC computer code has been developed for flow sheet simulation of pyrochemical processing of spent nuclear fuel. It utilizes a robust algorithm SLG for analyzing simultaneous chemical reactions between species distributed across many phases. Models have been developed for analysis of the oxide fuel reduction process, salt recovery by electrochemical decomposition of lithium oxide, uranium separation from the reduced fuel by electrorefining, and extraction of fission products into liquid cadmium. The versatility of GC is demonstrated by applying the code to a flow sheet of current interest

  2. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    1997-02-01

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

  3. Computer control of fuel handling activities at FFTF

    International Nuclear Information System (INIS)

    Romrell, D.M.

    1985-03-01

    The Fast Flux Test Facility near Richland, Washington, utilizes computer control for reactor refueling and other related core component handling and processing tasks. The computer controlled tasks described in this paper include core component transfers within the reactor vessel, core component transfers into and out of the reactor vessel, remote duct measurements of irradiated core components, remote duct cutting, and finally, transferring irradiated components out of the reactor containment building for off-site shipments or to long term storage. 3 refs., 16 figs

  4. 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)

  5. Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Wehner, T.R.

    1980-01-01

    The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

  6. Enhancing instruction in Fuels and Combustion Laboratory via a developed computer-assisted program for establishing efficient coal-diesel oil mixture (CDOM) fuel proportions

    Energy Technology Data Exchange (ETDEWEB)

    Maglaya, A.B. [La Salle University, Manila (Philippines). Dept. of Mechanical Engineering

    2004-07-01

    This paper discusses the relevance of digital computation in Fuels and Combustion Laboratory experiments used by the senior students of the Department of Mechanical Engineering, De La Salle University-Manila, Philippines. One of the students' experiments involved the determination of the most efficient CDOM fuel proportion as alternative fuel to diesel oil for steam generators and other industrial applications. Theoretical calculations show that it requires tedious and repetitive computations. A computer-assisted program was developed to lessen the time-consuming activities. The formulation of algorithms were based on the system of equations of the heat interaction between the CDOM fuel, combustion air and products of combustion and by applying the principles of mass and energy equations (or the First Law of Thermodynamics) for reacting systems were utilized. The developed computer-assisted program output verified alternative fuel selected through actual experimentation.

  7. Computer modelling of water reactor fuel element performance and life time

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Golovnin, I.S.; Elesin, V.F.

    1983-01-01

    Well calibrated models and methods of calculation permit the confident prediction of fuel element behaviour under most different operational conditions; based on the prediction of this kind one can improve designs and fuel element behaviour. Therefore, in the Soviet Union in the development of reactor cores for NPP one of the leading parts is given to design problems associated with computer modelling of fuel element performance and reliability. Special attention is paid to methods of calculation that permit the prediction of fuel element behaviour under conditions which either make experimental studies very complicated (practically impossible) or require laborious and expensive in-pile tests. Primarily it concerns accidents of different types, off-normal conditions, transients, fuel element behaviour at high burn-up, when an accumulation of a great amount of fission fragments is accompanied by changes in physical and mechanical properties as induced by irradiation damage, mechanical fatigue, physical and chemical reactions with a coolant, fission products etc. Some major computer modelling programs for the prediction of water reactor fuel behaviour are briefly described below and tendencies in the further development of work in this area are summarized

  8. Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

    International Nuclear Information System (INIS)

    Shimooke, Takanori; Yoshida, Kazuo

    1979-05-01

    The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

  9. Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature

    International Nuclear Information System (INIS)

    Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

    1982-10-01

    The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included

  10. FLATT - a computer programme for calculating flow and temperature transients in nuclear fuels

    International Nuclear Information System (INIS)

    Venkat Raj, V.; Koranne, S.M.

    1976-01-01

    FLATT is a computer code written in Fortran language for BESM-6 computer. The code calculates the flow transients in the coolant circuit of a nuclear reactor, caused by pump failure, and the consequent temperature transients in the fuel, clad, and the coolant. In addition any desired flow transient can be fed into the programme and the resulting temperature transients can be calculated. A case study is also presented. (author)

  11. Neutron Computed Tomography of Freeze/thaw Phenomena in Polymer Electrolyte Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Matthew M. Mech; Jack Brenizer; Kenan Unlu; A.K. Heller

    2008-12-12

    This report summarizes the final year's progress of the three-year NEER program. The overall objectives of this program were to 1) design and construct a sophisticated hight-resolution neutron computed tomography (NCT) facility, 2) develop novel and sophisticated liquid water and ice quantification analysis software for computed tomography, and 3) apply the advanced software and NCT capability to study liquid and ice distribution in polymer electrolyte fuel cells (PEFCs) under cold-start conditions. These objectives have been accomplished by the research team, enabling a new capability for advanced 3D image quantification with neutron imaging for fuel cell and other applications. The NCT water quantification methodology and software will greatly add to the capabilities of the neutron imaging community, and the quantified liquid water and ice distribution provided by its application to PEFCs will enhance understanding and guide design in the fuel cell community.

  12. Particle Image Velocimetry and Computational Fluid Dynamics Analysis of Fuel Cell Manifold

    DEFF Research Database (Denmark)

    Lebæk, Jesper; Blazniak Andreasen, Marcin; Andresen, Henrik Assenholm

    2010-01-01

    The inlet effect on the manifold flow in a fuel cell stack was investigated by means of numerical methods (computational fluid dynamics) and experimental methods (particle image velocimetry). At a simulated high current density situation the flow field was mapped on a 70 cell simulated cathode...

  13. CORCON: a computer program for modelling molten fuel/concrete interactions

    International Nuclear Information System (INIS)

    Muir, J.F.

    1980-01-01

    A computer program modelling the interaction between molten core materials and structural concrete is being developed to provide a capability for making quantitative estimates of reactor fuel-melt accidents. The principal phenomenological models, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. A code test comparison calculation is discussed

  14. Manual of phosphoric acid fuel cell power plant cost model and computer program

    Science.gov (United States)

    Lu, C. Y.; Alkasab, K. A.

    1984-01-01

    Cost analysis of phosphoric acid fuel cell power plant includes two parts: a method for estimation of system capital costs, and an economic analysis which determines the levelized annual cost of operating the system used in the capital cost estimation. A FORTRAN computer has been developed for this cost analysis.

  15. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

  16. Analysis of the Behavior of CAREM-25 Fuel Rods Using Computer Code BACO

    International Nuclear Information System (INIS)

    Estevez, Esteban; Markiewicz, Mario; Marino, Armando

    2000-01-01

    The thermo-mechanical behavior of a fuel rod subjected to irradiation is a complex process, on which a great quantity of interrelated physical-chemical phenomena are coupled.The code BACO simulates the thermo-mechanical behavior and the evolution of fission gases of a cylindrical rod in operation.The power history of fuel rods, arising from neutronic calculations, is the program input.The code calculates, among others, the temperature distribution and the principal stresses in the pellet and cladding, changes in the porosity and restructuring of pellet, the fission gases release, evolution of the internal gas pressure.In this work some of design limits of CAREM-25's fuel rods are analyzed by means of the computer code BACO.The main variables directly related with the integrity of the fuel rod are: Maximum temperature of pellet; Cladding hoop stresses; Gases pressure in the fuel rod; Cladding axial and radial strains, etc.The analysis of results indicates that, under normal operation conditions, the maximum fuel pellet temperature, cladding stresses, pressure of gases at end of life, etc, are below the design limits considered for the fuel rod of CAREM-25 reactor

  17. FUDA MOD-2: a computer program for simulation the performance of fuel element validation exercise

    International Nuclear Information System (INIS)

    Chouhan, S.K.; Tripathi, R.M.; Prasad, P.N.; Chauhan, Ashok

    2014-01-01

    The PHWR fuel element performance is evaluated using the fuel analysis computer code FUDA MOD2. It is specifically written for performance simulation of UO 2 fuel pellet, located in zirconium alloy sheath operating under coolant pressure. For specific element power histories, the code investigates the variables and their interactions that govern fuel element performance. The input data requires pellet dimensions, element dimensions, sheath properties, heat transfer data, thermal hydraulic parameters of coolant, the inner filler gas composition, flux gradient and linear heat ratings (LHR) at different burn up. The output data generated by the code are radial temperature profile of fuel and sheath, fuel sheath-gap heat transfer coefficient, fission gas generated and released, fission gas pressure, sheath stress and strain for different burn-up zones. The code has been verified against literature data and post irradiation examinations carried out. It is also bench marked against various international fuel element simulation programmes available with water cooled reactors operating countries. The present paper describes the FUDA MOD2 code verification studies carried out using the literature data and post irradiation examination data. (author)

  18. New weighted sum of gray gases model applicable to Computational Fluid Dynamics (CFD) modeling of oxy-fuel combustion

    DEFF Research Database (Denmark)

    Yin, Chungen; Johansen, Lars Christian Riis; Rosendahl, Lasse

    2010-01-01

    gases model (WSGGM) is derived, which is applicable to computational fluid dynamics (CFD) modeling of both air-fuel and oxy-fuel combustion. First, a computer code is developed to evaluate the emissivity of any gas mixture at any condition by using the exponential wide band model (EWBM...

  19. Computer programs for the in-core fuel management of power reactors

    International Nuclear Information System (INIS)

    1981-08-01

    This document gives a survey of the presently tested and used computer programs applicable to the in-core fuel management of light and heavy water moderated nuclear power reactors. Each computer program is described (provided that enough information was supplied) such that the nature of the physical problem solved and the basic mathematical or calculational approach are evident. In addition, further information regarding computer requirements, up-to-date applications and experiences and specific details concerning implementation, staff requirements, etc., are provided. Program procurement conditions, possible program implementation assistance and commercial conditions (where applicable) are given. (author)

  20. Development of a Computer Program for the Analysis Logistics of PWR Spent Fuels

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Choi, Jong Won; Cha, Jeong Hun

    2008-01-01

    It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  2. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

  3. Dynamics of nuclear fuel assemblies in vertical flow channels: computer modelling and associated studies

    International Nuclear Information System (INIS)

    Mason, V.A.; Pettigrew, M.J.; Lelli, G.; Kates, L.; Reimer, E.

    1978-10-01

    A computer model, designed to predict the dynamic behaviour of nuclear fuel assemblies in axial flow, is described in this report. The numerical methods used to construct and solve the matrix equations of motion in the model are discussed together with an outline of the method used to interpret the fuel assembly stability data. The mathematics developed for forced response calculations are described in detail. Certain structural and hydrodynamic modelling parameters must be determined by experiment. These parameters are identified and the methods used for their evaluation are briefly described. Examples of typical applications of the dynamic model are presented towards the end of the report. (author)

  4. CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1989-02-01

    A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  5. FRAP-T1: a computer code for the transient analysis of oxide fuel rods

    International Nuclear Information System (INIS)

    Dearien, J.A.; Miller, R.L.; Hobbins, R.R.; Siefken, L.J.; Baston, V.F.; Coleman, D.R.

    1977-02-01

    FRAP-T is a FORTRAN IV computer code which can be used to solve for the transient response of a light water reactor (LWR) fuel rod during accident transients such as loss-of-coolant accident (LOCA) or a power-cooling-mismatch (PCM). The coupled effects of mechanical, thermal, internal gas, and material property response on the behavior of the fuel rod are considered. FRAP-T is a modular code with each major computational model isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAP-T is coupled to a materials properties subcode (MATPRO) which is used to provide gas, fuel, and cladding properties to the FRAP-T computational subcodes. No material properties need be supplied by the code user. The needed water properties are stored in tables built into the code. Critical heat flux (CHF) and heat transfer correlations for a wide range of coolant conditions are contained in modular subroutines. FRAP-T has been evaluated by making extensive comparisons between predictions of the code and experimental data. Comparison of predicted and experimental results are presented for a range of FRAP-T calculated parameters. The code is presently programmed and running on an IBM-360/75 and a CDC 7600 computer

  6. Application of a personal computer relational data base management system to fuel cycle economic scoping

    International Nuclear Information System (INIS)

    Malone, J.P.; Dooley, G.D.

    1986-01-01

    A personal computer (PC) relational data base management system (RDBMS) permits large quantities of data to be maintained in a data base composed of structured data sets or files and provides data access through a software environment, procedure, or program language. The features of an RDBMS-based system create an environment on a PC that can provide significant benefits to any fuel cycle economics analysis. The ability to maintain a separate data set for each fuel cycle parameter group and the ability to manipulate the data through a series of independent calculation modules combine to provide the fuel cycle analyst with more time to examine and use the data, because less time is required to manipulate it

  7. A computer model to predict temperatures and gas flows during AGR fuel handling

    International Nuclear Information System (INIS)

    Bishop, D.C.; Bowler, P.G.

    1986-01-01

    The paper describes the development of a comprehensive computer model (HOSTAGE) that has been developed for the Heysham II/Torness AGRs to predict temperature transients for all the important components during normal and fault conditions. It models not only the charge and discharge or fuel from an on-load reactor but also follows the fuel down the rest of the fuel route until it is dismantled. The main features of the physical model of gas and heat flow are described. Experimental results are used where appropriate and an indication will be given of how the predictions by HOSTAGE correlate with operating AGR reactors. The role of HOSTAGE in the Heysham II/Torness safety case is briefly discussed. (author)

  8. SIEX3: A correlated computer code for prediction of fast reactor mixed oxide fuel and blanket pin performance

    International Nuclear Information System (INIS)

    Baker, R.B.; Wilson, D.R.

    1986-04-01

    The SIEX3 computer program was developed to calculate the fuel and cladding performance of oxide fuel and oxide blanket pins irradiated in the fast neutron environment of a liquid metal cooled reactor. The code is uniquely designed to be accurate yet quick running and use a minimum of computer core storage. This was accomplished through the correlation of physically based models to very large data bases of irradiation test results. Data from over 200 fuel pins and over 800 transverse fuel microscopy samples were used in the calibrations

  9. Computer modelling of fuel behaviour and performance in research reactors as related to fuel depletion

    International Nuclear Information System (INIS)

    Akaho, E.H.K.; Danso, K.A.

    1990-07-01

    The code ISODEP is being developed to compute the rate of production of nuclides in different homogeneous zones of a research reactor. ISODEP will form part of a main programme MEPE for analysis of the Miniature Neutron Source Reactor. An exponential method proposed by Hansen is the basis for the numerical solution of non-homogeneous simultaneous equations which describe the rate of production and decay of nuclides. Four principal chains associated with uranium-fuelled reactor were studied. The trend of results was found to be consistent with those obtained by analytical method. It is hoped that after slight modifications of the code and with appropriate effective microscopic cross-section data it will be suitable for research reactor analysis. (author)

  10. COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

    International Nuclear Information System (INIS)

    Rhodes, C.A.

    1984-12-01

    This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

  11. A computational model of a PEM fuel cell with finite vapor absorption rate

    Energy Technology Data Exchange (ETDEWEB)

    Vorobev, A.; Zikanov, O.; Shamim, T. [Department of Mechanical Engineering, University of Michigan-Dearborn, 48128-1491 Dearborn, MI (United States)

    2007-03-30

    The paper presents a new computational model of non-steady operation of a PEM fuel cell. The model is based on the macroscopic hydrodynamic approach and assumptions of low humidity operation and one-dimensionality of transport processes. Its novelty and advantage in comparison with similar existing models is that it takes into account the finite-time equilibration between vapor and membrane-phase liquid water within the catalyst layers. The phenomenon is described using an additional parameter with the physical meaning of the typical reciprocal time of the equilibration. A computational parametric study is conducted to identify the effect of the finite-time equilibration on steady-state and transient operation of a PEM fuel cell. (author)

  12. Integration of computational modeling and experimental techniques to design fuel surrogates

    DEFF Research Database (Denmark)

    Choudhury, H.A.; Intikhab, S.; Kalakul, Sawitree

    2017-01-01

    performance. A simplified alternative is to develop surrogate fuels that have fewer compounds and emulate certain important desired physical properties of the target fuels. Six gasoline blends were formulated through a computer aided model based technique “Mixed Integer Non-Linear Programming” (MINLP...... Virtual Process-Product Design Laboratory (VPPD-Lab) are applied onto the defined compositions of the surrogate gasoline. The aim is to primarily verify the defined composition of gasoline by means of VPPD-Lab. ρ, η and RVP are calculated with more accuracy and constraints such as distillation curve...... and flash point on the blend design are also considered. A post-design experiment-based verification step is proposed to further improve and fine-tune the “best” selected gasoline blends following the computation work. Here, advanced experimental techniques are used to measure the RVP, ρ, η, RON...

  13. Evaluation of pulmonary changes due to biomass fuels using high-resolution computed tomography

    International Nuclear Information System (INIS)

    Kara, Mustafa; Tas, Fikret; Bulut, Sema; Akkurt, Ibrahim; Seyfikli, Zehra

    2003-01-01

    Biomass fuels are frequently used in rural areas of the world for cooking and heating frequently. It has been reported that the use of these fuels causes hazardous effects on the lungs. In this study, we evaluated the pulmonary changes due to the use of biomass fuels in a female population that lives in our territory by high-resolution computed tomography (HRCT). The study analyzed three groups of women. The first group comprised those subjects who were exposed to biomass without respiratory symptoms (group 1; n=32). The second group comprised those individuals that were exposed to biomass and showed respiratory symptoms, such as cough, sputum production, and dyspnea (group 2; n=30). The third group was composed of women who were not exposed to biomass and also had no respiratory symptoms (group 3; n=30). Women with a history of concomitant pulmonary diseases were excluded from the study. All groups were examined with HRCT. Groups 1 and 2 (individuals exposed to biomass fuels) had more pathologic findings than group 3 (not exposed to biomass fuels). Ground-glass appearance was seen in 71.9% in group 1, 23.3% in group 2, and 3.3% in group 3. The difference between the groups was statistically significant (p<0.05). Fibrotic bands were seen 50% in group 1, 63.3% in group 2, and only 6.7% in group 3 (p<0.001). Exposure to biomass fuels was the cause or predisposing factor for many pulmonary diseases, ranging from chronic bronchitis to diffuse lung diseases. We believe that these pathological changes due to biomass fuels can be detected earlier by HRCT and the diseases might be prevented or treated earlier. (orig.)

  14. Citham a computer code for calculating fuel depletion-description, tests, modifications and evaluation

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1984-12-01

    The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author) [pt

  15. Neutronics investigation of CANDU deuterium uranium 6 reactor fueled (transuranic-TH) O-2 using a computational method

    Energy Technology Data Exchange (ETDEWEB)

    Gholamzadeh, Zohreh; Mirvakili, Seyed Mohammad; Khalafi, Hossein [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

    2015-02-15

    241Am, 243Am, and 237Np isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

  16. Development of fuel prices and its impact on the future development of nuclear energy, the use of computer code DESAE

    International Nuclear Information System (INIS)

    Panik, M.; Necas, V.

    2007-01-01

    The thesis is an overview of fuel prices, its key components, such as the particular price and price of natural uranium fuel enrichment. The paper outlines the expected impact of higher fuel prices on the future development of nuclear energy. The last section is devoted to computer code DESAE, designed to calculate and compare advantages and disadvantages of different nuclear systems, but also to calculate the parameters of given nuclear system. They suggested the possibility of using code in practice. (author)

  17. KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

    International Nuclear Information System (INIS)

    Inhoven, H.

    1984-12-01

    KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.) [de

  18. Development of computer models for fuel element behaviour in water reactors

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1987-03-01

    Description of fuel behaviour during normal operation transients and accident conditions has always represented a most challenging and important problem. Reliable predictions constitute a basic demand for safety based calculations, for design purposes and for fuel performance. Therefore, computer codes based on deterministic and probabilistic models were developed. Possibility of comprehensive descriptions of the phenomena is precluded in view of the great number of individual processes, involving physical, chemical, thermohydraulical and mechanical parameters, to be considered in a wide range of situations. In case of fast thermal transients predictive capability is limited by the kinetics of evolution of the system and its eventual dynamic behaviour. Evidently, probabilistic approaches are also limited by the sparcity and limited breadth of the impirical data base. Code predictions have to be evaluated against power reactor data, results from simulation experiments and, if possible, include cross validation of different codes and validation of sub-models. Progress on this subject is reviewed in this report, which completes the co-ordinated research programme on 'Development of Computer Models for Fuel Element Behaviour in Water Reactors' (D-COM), initiated under the auspices of the IAEA in 1981

  19. Computational fluid dynamics application: slosh analysis of a fuel tank model

    International Nuclear Information System (INIS)

    Iu, H.S.; Cleghorn, W.L.; Mills, J.K.

    2004-01-01

    This paper presents the analysis of fluid slosh behaviour inside a fuel tank model. The fuel tank model was a simplified version of a stock fuel tank that has a sloshing noise problem. A commercial CFD software, FLOW-3D, was used to simulate the slosh behaviour. Slosh experiments were performed to verify the computer simulation results. High speed video equipment enhanced with a data acquisition system was used to record the slosh experiments and to obtain the instantaneous sound level of each video frame. Five baffle configurations including the no baffle configuration were considered in the computer simulations and the experiments. The simulation results showed that the best baffle configuration can reduce the mean kinetic energy by 80% from the no baffle configuration in a certain slosh situation. The experimental results showed that 15dB(A) noise reduction can be achieved by the best baffle configuration. The correlation analysis between the mean kinetic energy and the noise level showed that high mean kinetic energy of the fluid does not always correspond to high sloshing noise. High correlation between them only occurs for the slosh situations where the fluid hits the top of the tank and creates noise. (author)

  20. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation

    International Nuclear Information System (INIS)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

    2015-01-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  1. Modification in the FUDA computer code to predict fuel performance at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Das, M; Arunakumar, B V; Prasad, P N [Nuclear Power Corp., Mumbai (India)

    1997-08-01

    The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig.

  2. Modification in the FUDA computer code to predict fuel performance at high burnup

    International Nuclear Information System (INIS)

    Das, M.; Arunakumar, B.V.; Prasad, P.N.

    1997-01-01

    The computer code FUDA (FUel Design Analysis) participated in the blind exercises organized by the IAEA CRP (Co-ordinated Research Programme) on FUMEX (Fuel Modelling at Extended Burnup). While the code prediction compared well with the experiments at Halden under various parametric and operating conditions, the fission gas release and fission gas pressure were found to be slightly over-predicted, particularly at high burnups. In view of the results of 6 FUMEX cases, the main models and submodels of the code were reviewed and necessary improvements were made. The new version of the code FUDA MOD 2 is now able to predict fuel performance parameter for burn-ups up to 50000 MWD/TeU. The validation field of the code has been extended to prediction of thorium oxide fuel performance. An analysis of local deformations at pellet interfaces and near the end caps is carried out considering the hourglassing of the pellet by finite element technique. (author). 15 refs, 1 fig

  3. Calculation of Plutonium content in RSG-GAS spent fuel using IAFUEL computer code

    International Nuclear Information System (INIS)

    Mochamad-Imron

    2003-01-01

    It has been calculated the contain of isotopes Pu-239, Pu-240, Pu-241, and isotope Pu-242 in MTR reactor fuel types which have U-235 contain about 250 gram. The calculation was performed in three steps. The first step is to determine the library of calculation output of BOC (Beginning of Cycle). The second step is to determine the core isotope density, the weight of plutonium for one core, and one fuel isotope density. The third step is to calculate weight of plutonium in gram. All calculation is performed by IAFUEL computer code. The calculation was produced content of each Pu isotopes were Pu-239 is 6.7666 gr, Pu-240 is 1.4628 gr, Pu-241 is 0.52951 gr, and Pu-242 is 0.068952 gr

  4. Computational Thermodynamics Analysis of Vaporizing Fuel Droplets in the Human Upper Airways

    Science.gov (United States)

    Zhang, Zhe; Kleinstreuer, Clement

    The detailed knowledge of air flow structures as well as particle transport and deposition in the human lung for typical inhalation flow rates is an important precursor for dosimetry-and-health-effect studies of toxic particles as well as for targeted drug delivery of therapeutic aerosols. Focusing on highly toxic JP-8 fuel aerosols, 3-D airflow and fluid-particle thermodynamics in a human upper airway model starting from mouth to Generation G3 (G0 is the trachea) are simulated using a user-enhanced and experimentally validated finite-volume code. The temperature distributions and their effects on airflow structures, fuel vapor deposition and droplet motion/evaporation are discussed. The computational results show that the thermal effect on vapor deposition is minor, but it may greatly affect droplet deposition in human airways.

  5. Computational fluid dynamics (CFD) modelling of coal/biomass co-firing in pulverised fuel boilers

    Energy Technology Data Exchange (ETDEWEB)

    Moghtaderi, B.; Meesri, C. [University of Newcastle, Callaghan, NSW (Australia). CRC for Coal in Sustainable Development, Dept. of Chemical Engineering

    2002-07-01

    The present study is concerned with computational fluid dynamics (CFD) modelling of coal/biomass blends co-fired under conditions pertinent to pulverised fuel (PF) boilers. The attention is particularly focused on the near burner zone to examine the impact of biomass on the flame geometry and temperature. The predictions are obtained by numerical solution of the conservation equations for the gas and particle phases. The gas phase is solved in the Eulerian domain using steady-state time-averaged Navier-Stokes equations while the solution of the particle phase is obtained from a series of Lagrangian particle tracking equations. Turbulence is modelled using the {kappa}-{epsilon} and Reynolds Stress models. The comparison between the predictions and experimental measurement reported in the literature resulted in a good agreement. Other influences of biomass co-firing are observed for fuel devolatilisation and burnout. 19 refs., 6 figs.

  6. Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

    International Nuclear Information System (INIS)

    2007-12-01

    Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element

  7. Accuracy and computational time of a hierarchy of growth rate definitions for breeder reactor fuel

    International Nuclear Information System (INIS)

    Maudlin, P.J.; Borg, R.C.; Ott, K.O.

    1979-01-01

    For a hierarchy of four logically different definitions for calculating the asymptotic growth of fast breeder reactor fuel, an investigation is performed concerning the comparative accuracy and computational effort associated with each definition. The definition based on detailed calculation of the accumulating fuel in an expanding park of reactors asymptotically yields the most accurate value of the infinite time growth rate, γ/sup infinity/, which is used as a reference value. The computational effort involved with the park definition is very large. The definition based on the single reactor calculation of the equilibrium surplus production rate and fuel inventory gives a value for γ/sup infinity of comparable accuracy to the park definition and uses significantly less central processor unit (CPU) time. The third definition is based on a continuous treatment of the reactor fuel cycle for a single reactor and gives a value for γ/sup infinity/ that accurately approximates the second definition. The continuous definition requires very little CPU time. The fourth definition employs the isotopic breeding worths, w/sub i//sup */, for a projection of the asymptotic growth rate. The CPU time involved in this definition is practically nil if its calculation is based on the few-cycle depletion calculation normally performed for core design and critical enrichment evaluations. The small inaccuracy (approx. = 1%) of the breeding-worth-based definition is well within the inaccuracy range that results unavoidably from other sources such as nuclear cross sections, group constants, and flux calculations. This fully justifies the use of this approach in routine calculations

  8. Computational Flame Diagnostics for Direct Numerical Simulations with Detailed Chemistry of Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Tianfeng [Univ. of Connecticut, Storrs, CT (United States)

    2017-02-16

    The goal of the proposed research is to create computational flame diagnostics (CFLD) that are rigorous numerical algorithms for systematic detection of critical flame features, such as ignition, extinction, and premixed and non-premixed flamelets, and to understand the underlying physicochemical processes controlling limit flame phenomena, flame stabilization, turbulence-chemistry interactions and pollutant emissions etc. The goal has been accomplished through an integrated effort on mechanism reduction, direct numerical simulations (DNS) of flames at engine conditions and a variety of turbulent flames with transport fuels, computational diagnostics, turbulence modeling, and DNS data mining and data reduction. The computational diagnostics are primarily based on the chemical explosive mode analysis (CEMA) and a recently developed bifurcation analysis using datasets from first-principle simulations of 0-D reactors, 1-D laminar flames, and 2-D and 3-D DNS (collaboration with J.H. Chen and S. Som at Argonne, and C.S. Yoo at UNIST). Non-stiff reduced mechanisms for transportation fuels amenable for 3-D DNS are developed through graph-based methods and timescale analysis. The flame structures, stabilization mechanisms, local ignition and extinction etc., and the rate controlling chemical processes are unambiguously identified through CFLD. CEMA is further employed to segment complex turbulent flames based on the critical flame features, such as premixed reaction fronts, and to enable zone-adaptive turbulent combustion modeling.

  9. FRAPCON-2: A Computer Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A; Bohn, M. P.; Rausch, W. N.; Williford, R. E.; Lanning, D. D.

    1981-01-01

    FRAPCON-2 is a FORTRAN IV computer code that calculates the steady state response of light Mater reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, deformation, and tai lure histories of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (a) heat conduction through the fuel and cladding, (b) cladding elastic and plastic deformation, (c) fuel-cladding mechanical interaction, (d) fission gas release, (e} fuel rod internal gas pressure, (f) heat transfer between fuel and cladding, (g) cladding oxidation, and (h) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat transfer correlations. FRAPCON-2 is programmed for use on the CDC Cyber 175 and 176 computers. The FRAPCON-2 code Is designed to generate initial conditions for transient fuel rod analysis by either the FRAP-T6 computer code or the thermal-hydraulic code, RELAP4/MOD7 Version 2.

  10. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

    2005-01-01

    Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

  11. Application of Computational Intelligence Methods to In-Core Fuel Management

    International Nuclear Information System (INIS)

    Erdogan, A.

    2001-01-01

    In this study, a computer program package has been developed which supports the in-core fuel management activities for pressurized water reactors, generates and recommends an optimum loading pattern to ensure safe and efficient reactor operation. A search for an optimum fuel loading pattern must be conducted in the space of several core parameters such as power distribution, which is an excessively time consuming computational process. Global core calculation codes take a relatively long time to do the task. The time interval necessary for the iterative process was reduced by using an artificial neural network estimator for the calculations. In this way, it was possible to analyze more loading patterns in the same time interval and the probability of finding a desired optimum was increased. As a case study, the core of the Almaraz Nuclear Plant of Spain, a pressurized water reactor, was modeled for the core calculation code system. The 2-group cross sections for the fuel assembly types were calculated and stored for later usage with the diffusion code. 2000 loading patterns were generated by placing fuel assemblies to random positions in the core, and for each pattern the power distribution and effective multiplication factor (k e ff) were calculated with the diffusion code. At the next stage, 500 of the loading patterns were introduced to the neural network as input data for the training process. The remaining 1500 patterns were used to validate the neural network implementation. It was shown that the neural network estimates the power distribution and the K effective within acceptable error limits. To complete the system, a loading pattern generator was developed. This module consists of a set of rules and an algorithm that places the fuel assemblies to core positions. The neural network estimated the power distribution and k e ff for the loading patterns that were generated by this module. The patterns that have a maximum power fraction lower than, and a minimum

  12. URANUS - a computer programme for the thermal and mechanical analysis of the fuel rods in a nuclear reactor

    International Nuclear Information System (INIS)

    Lassmann, K.

    1978-01-01

    The URANUS code, a digital computer programme for the thermal and mechanical analysis of integral fuel rods, is described. With this code the fuel rods found in the majority of power reactors can be analyzed. URANUS is built around a quasi two-dimensional analysis of fuel and cladding. The mechanical analysis can accommodate seven components of strain: elastic, time-independent plastic, creep and thermal strains, as well as strains due to swelling, cracking and densification. The heat generation and temperature distribution, cladding/fuel gap closure, pellet cracking and crack healing, fission-gas release, corrosion, O/M-distribution and plutonium redistribution are modelled. Geometric non-linearities (large displacements) are included; steady state or transient loading (pressure, temperature) is possible. In this paper special attention is paid to a theory for determining crack structures. The present status of the URANUS computer programme and a critical comparison with other fuel rod codes as well as sample analyses are given. (Auth.)

  13. EXPEL - a computing module for molten fuel/coolant interactions in fast reactor sub-assemblies

    International Nuclear Information System (INIS)

    Fishlock, T.P.

    1975-10-01

    This report describes a module for computing the effects of a molten fuel/coolant interaction in a fast reactor subassembly. The module is to be incorporated into the FRAX code which calculates the consequences of hypothetical whole core accidents. Details of the interaction are unknown and in consequence the model contains a large number of parameters which must be set by assumption. By variation of these parameters the interaction may be made mild or explosive. Results of a parametric survey are included. (author)

  14. FRAP-T6: a computer code for the transient analysis of oxide fuel rods

    International Nuclear Information System (INIS)

    Siefken, L.J.; Shah, V.N.; Berna, G.A.; Hohorst, J.K.

    1983-06-01

    FRAP-T6 is a computer code which is being developed to calculate the transient behavior of a light water reactor fuel rod. This report is an addendum to the FRAP-T6/MODO user's manual which provides the additional user information needed to use FRAP-T6/MOD1. This includes model changes, improvements, and additions, coding changes and improvements, change in input and control language, and example problem solutions to aid the user. This information is designed to supplement the FRAP-T6/MODO user's manual

  15. Computational Investigation of On-Line Interrogation of Pebble Bed Reactor Fuel

    Science.gov (United States)

    Hawari, A. I.; Chen, Jianwei

    2005-10-01

    Pebble bed reactors are characterized by multipass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit (80000-100000 MWD/MTU). For such reactors, the fuel is assayed on-line to ensure that the burnup limit is not breached. We considered assaying the fuel using an HPGe detector to perform passive gamma-ray spectrometry of fission products. Since neither fresh nor irradiated fuel is readily available, computer simulations were utilized to identify the radionuclides that can be used as burnup indicators, and to visualize the gamma-ray spectra at various levels of burnup. Specifically, we used the ORIGEN-MONTEBURNS-MCNP code system. This allowed the establishment of the burnup dependent one-group gas reactor cross-sections for the radionuclides of interest. Subsequently, ORIGEN was used to simulate in-core pebble depletion to establish the irradiated pebble isotopics. Finally, the codes MCNP and SYNTH were used to simulate the response of the HPGe gamma-ray spectrometer. The results show that absolute and relative indicators can be used on-line to determine unambiguously the enrichment and burnup on a pebble-by-pebble basis. The activity of Cs-137 or the activity ratio of Co-60/Cs-134 can be combined with the activity ratio of Np-239/I-132 to yield the enrichment and burnup information. To use the relative indicators, a relative efficiency calibration of the gamma-ray spectrometer can be performed using the La-140 gamma lines that are emitted by the irradiated pebble. I-132, Cs-134, Cs-137, La-140, and Np-239 are produced upon the irradiation of the fuel. Co-60 is produced by doping the fuel with a small amount (/spl sim/100 ppm) of Co-59. Using this approach, the uncertainty in burnup determination due to factors such as power history variation, detector efficiency calibration, and counting statistics is expected to remain in the range of /spl plusmn/5% to /spl plusmn/10%.

  16. Improvement of the computing speed of the FBR fuel pin bundle deformation analysis code 'BAMBOO'

    International Nuclear Information System (INIS)

    Ito, Masahiro; Uwaba, Tomoyuki

    2005-04-01

    JNC has developed a coupled analysis system of a fuel pin bundle deformation analysis code 'BAMBOO' and a thermal hydraulics analysis code ASFRE-IV' for the purpose of evaluating the integrity of a subassembly under the BDI condition. This coupled analysis took much computation time because it needs convergent calculations to obtain numerically stationary solutions for thermal and mechanical behaviors. We improved the computation time of the BAMBOO code analysis to make the coupled analysis practicable. 'BAMBOO' is a FEM code and as such its matrix calculations consume large memory area to temporarily stores intermediate results in the solution of simultaneous linear equations. The code used the Hard Disk Drive (HDD) for the virtual memory area to save Random Access Memory (RAM) of the computer. However, the use of the HDD increased the computation time because Input/Output (I/O) processing with the HDD took much time in data accesses. We improved the code in order that it could conduct I/O processing only with the RAM in matrix calculations and run with in high-performance computers. This improvement considerably increased the CPU occupation rate during the simulation and reduced the total simulation time of the BAMBOO code to about one-seventh of that before the improvement. (author)

  17. Computer code for the analysis of destructive pressure generation process during a fuel failure accident, PULSE-2

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1978-03-01

    The computer code PULSE-2 has been developed for the analysis of pressure pulse generation process when hot fuel particles come into contact with the coolant in a fuel rod failure accident. In the program, it is assumed that hot fuel fragments mix with the coolant instantly and homogeneously in the failure region. Then, the rapid vaporization of the coolant and transient pressure rise in failure region, and the movement of ejected coolant slugs are calculated. The effect of a fuel-particle size distribution is taken into consideration. Heat conduction in the fuel particles and heat transfer at fuel-coolant interface are calculated. Temperature, pressure and void fraction in the mixed region are calculated from the average enthalpy. With physical property subroutines for liquid sodium and water, the model is usable for both LMFBR and LWR conditions. (auth.)

  18. Development of COMPAS, computer aided process flowsheet design and analysis system of nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Homma, Shunji; Sakamoto, Susumu; Takanashi, Mitsuhiro; Nammo, Akihiko; Satoh, Yoshihiro; Soejima, Takayuki; Koga, Jiro; Matsumoto, Shiro

    1995-01-01

    A computer aided process flowsheet design and analysis system, COMPAS has been developed in order to carry out the flowsheet calculation on the process flow diagram of nuclear fuel reprocessing. All of equipments, such as dissolver, mixer-settler, and so on, in the process flowsheet diagram are graphically visualized as icon on a bitmap display of UNIX workstation. Drawing of a flowsheet can be carried out easily by the mouse operation. Not only a published numerical simulation code but also a user's original one can be used on the COMPAS. Specifications of the equipment and the concentration of components in the stream displayed as tables can be edited by a computer user. Results of calculation can be also displayed graphically. Two examples show that the COMPAS is applicable to decide operating conditions of Purex process and to analyze extraction behavior in a mixer-settler extractor. (author)

  19. Theory of the space-dependent fuel management computer code ''UAFCC''

    International Nuclear Information System (INIS)

    El-Meshad, Y.; Morsy, S.; El-Osery, I.A.

    1981-01-01

    This report displays the theory of the spatial burnup computer code ''UAFCC'' which has been constructed as a part of an integrated reactor calculation scheme proposed at the Reactors Department of the ARE Atomic Energy Authority. The ''UAFCC'' is a single energy-one-dimensional diffusion burnup FORTRAN computer code for well moderated, multiregion, cylindrical thermal reactors. The effect of reactivity variation with burnup is introduced in the steady state diffusion equation by a fictitious neutron source. The infinite multiplication factor, the total migration area, and the power density per unit thermal flux are calculated from the point model burnup code ''UABUC'' fitted to polynomials of suitable degree in the flux-time, and then used as an input data to the ''UAFCC'' code. The proposed burnup spatial model has been used to study the different stratogemes of the incore fuel management schemes. The conclusions of this study will be presented in a future publication. (author)

  20. Personal computer based decision support system for routing nuclear spent fuel

    International Nuclear Information System (INIS)

    Chin, Shih-Miao; Joy, D.S.; Johnson, P.E.; Bobic, S.M.; Miaou, Shaw-Pin

    1989-01-01

    An approach has been formulated to route nuclear spent fuel over the US Interstate highway network. This approach involves the generation of alternative routes so that any potential adverse impacts will not only concentrate on regions along the shortest path between the nuclear power plant and repository. Extensive literature research on the shortest path finding algorithms has been carried out. Consequently, an extremely efficient shortest path algorithm has been implemented and significantly increases the overall system performance. State-of-the-art interactive computer graphics is used. In addition to easy-to-use pop-up menus, full color mapping and display capabilities are also incorporated. All of these features have been implemented on commonly available personal computers. 6 figs., 2 tabs

  1. Comparison of structural computer programs used for the analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Friley, J.R.

    1984-09-01

    Several structural analysis computer programs were selected and used in analyses relevant to the hypothetical impact requirements for spent fuel shipping cask designs. The objective of the study was to evaluate the computer codes by performing a series of analyses and comparing results. The code evaluation efforts treated end and side impact situations only. As a result, the models were either one or two dimensional. Both clad lead and solid wall construction types were considered. For clad lead models, frictionless sliding between the lead and cladding was assumed. General agreement was achieved between the codes for problems involving non-clad models. For clad models, agreement between the codes was poor. This work was sponsored by the Department of Energy through the Transportation Technology Center at Sandia National Laboratories, Albuquerque, New Mexico. 16 references, 18 figures, 9 tables

  2. RAP-3A Computer code for thermal and hydraulic calculations in steady state conditions for fuel element clusters

    International Nuclear Information System (INIS)

    Popescu, C.; Biro, L.; Iftode, I.; Turcu, I.

    1975-10-01

    The RAP-3A computer code is designed for calculating the main steady state thermo-hydraulic parameters of multirod fuel clusters with liquid metal cooling. The programme provides a double accuracy computation of temperatures and axial enthalpy distributions of pressure losses and axial heat flux distributions in fuel clusters before boiling conditions occur. Physical and mathematical models as well as a sample problem are presented. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

  3. DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

    Energy Technology Data Exchange (ETDEWEB)

    Genin, J.B. [DEN/DEC/S3C, CEA Cadarache, 13 - Saint-Paul-lez-Durance (France); Harrer, A. [EdF/SEPTEN, 69 - Villeurbanne (France); Musante, Y. [FRAMATOME-ANP, 69 - Lyon (France)

    2002-07-01

    DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

  4. DIADEME: A computer code to assess in operation defective fuel characteristics and primary circuit contamination

    International Nuclear Information System (INIS)

    Genin, J.B.; Harrer, A.; Musante, Y.

    2002-01-01

    DIADEME is a computer code developed within the framework of R and D cooperation between the French Atomic Energy Commission (CEA), Electricite de France (EdF) and FRAMATOME-ANP. Its aim is to assess in operation defective fuel characteristics and primary circuit contamination for actinides and long half-life fission products involved in health physics problems as well as in waste and decommissioning studies. DIADEME has been developed and qualified for the EDF nuclear power plants. For many years, both theoretical and experimental studies have been carried out at the CEA on the release of fission products and actinides out of defective fuel rods in operation, their migration and deposition in PWR primary circuits. These studies have allowed defect characteristic diagnosis methods to be developed, based on radiochemical measurements of the primary coolant. These methods are generally used along with gamma spectrometry measurements on primary water sampling. In order to be completely efficient, these methods can also be used in connection with an on-line primary water gamma spectrometry device. This permits to obtain the most comprehensive data on fission product activity evolutions at steady state and during operation transients, and allows the on-line characterization of the defective fuel assemblies. For long half-life fission products and for actinides, DIADEME is also able to assess the activities of soluble and insoluble forms in the primary water and in the chemical and voluminal control system (CVCS) filters and resins, as well as those activities deposited on primary circuit surfaces. (author)

  5. Computational simulation of water transport in PEM fuel cells using an improved membrane model

    International Nuclear Information System (INIS)

    Cao, J.; Djilali, N.

    2000-01-01

    Computational models and simulation tools can provide valuable insight and guidance for design, performance optimization, and cost reduction of fuel cells. In proton-exchange membrane fuel cells it is particularly important to maintain appropriate water content and temperature in the electrolyte membrane. In this paper we describe a mathematical model for the membrane that takes into account the diffusion of water, the pressure variation, and the electro-osmotic drag in the membrane. Applying conservation laws for water and current and using an empirical relationship between electro-osmotic drag and water content, we obtain a transport equation for water molar concentration and derive a new equation for the electric potential that accounts for variable water content and is more accurate than the conventionally employed Laplace's equation does. The model is coupled with a computational fluid dynamics model for diffusive transport in the electrodes and convective transport in the reactant flow channels. Simulations for a two-dimensional cell are performed over nominal current densities ranging form i=0.1 A/cm≅ to 1.2 A/cm≅. The impact and importance of temperature and pressure non-uniformity, and of two-dimensionality are assessed and discussed. (author)

  6. Investigation on structural analysis computer program of spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods such as HONDO, STEALTH and DYNA-3D, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for 9 m vertical drop impact of the axisymmetric lead cylinder with and without stainless steel clads were adopted where the calculational evaluations for taking into account the strain rate effect were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  7. Investigation on structural analysis computer program of spent nuclear fuel shipping cask, (2)

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results (II) done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for drop impact of the axisymmetric cylinder and plate were adopted the calculational evaluations for taking into account the strain rate effect of material properties, effect of artificial viscosity and effect of time integration step size were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  8. GAPCON-THERMAL-2: a computer program for calculating the thermal behavior of an oxide fuel rod

    International Nuclear Information System (INIS)

    Beyer, C.E.; Hann, C.R.; Lanning, D.D.; Panisko, F.E.; Parchen, L.J.

    1975-11-01

    A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON-THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady-state thermal performance and to provide input for accident analyses. Some models used in the code provide best estimate as well as conservative predictions. Each of the individual models in the code is based on the best available data

  9. Neutronics investigation of CANada Deuterium Uranium 6 reactor fueled (transuranic–Th) O2 using a computational method

    OpenAIRE

    Zohreh Gholamzadeh; Seyed Mohammad Mirvakili; Hossein Khalafi

    2015-01-01

    Background: 241Am, 243Am, and 237Np isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic b...

  10. Computational Methods for Nanoscale X-ray Computed Tomography Image Analysis of Fuel Cell and Battery Materials

    Science.gov (United States)

    Kumar, Arjun S.

    Over the last fifteen years, there has been a rapid growth in the use of high resolution X-ray computed tomography (HRXCT) imaging in material science applications. We use it at nanoscale resolutions up to 50 nm (nano-CT) for key research problems in large scale operation of polymer electrolyte membrane fuel cells (PEMFC) and lithium-ion (Li-ion) batteries in automotive applications. PEMFC are clean energy sources that electrochemically react with hydrogen gas to produce water and electricity. To reduce their costs, capturing their electrode nanostructure has become significant in modeling and optimizing their performance. For Li-ion batteries, a key challenge in increasing their scope for the automotive industry is Li metal dendrite growth. Li dendrites are structures of lithium with 100 nm features of interest that can grow chaotically within a battery and eventually lead to a short-circuit. HRXCT imaging is an effective diagnostics tool for such applications as it is a non-destructive method of capturing the 3D internal X-ray absorption coefficient of materials from a large series of 2D X-ray projections. Despite a recent push to use HRXCT for quantitative information on material samples, there is a relative dearth of computational tools in nano-CT image processing and analysis. Hence, we focus on developing computational methods for nano-CT image analysis of fuel cell and battery materials as required by the limitations in material samples and the imaging environment. The first problem we address is the segmentation of nano-CT Zernike phase contrast images. Nano-CT instruments are equipped with Zernike phase contrast optics to distinguish materials with a low difference in X-ray absorption coefficient by phase shifting the X-ray wave that is not diffracted by the sample. However, it creates image artifacts that hinder the use of traditional image segmentation techniques. To restore such images, we setup an inverse problem by modeling the X-ray phase contrast

  11. FEMAXI-III: a computer code for the analysis of thermal and mechanical behavior of fuel rods

    International Nuclear Information System (INIS)

    Nakajima, Tetsuo; Ichikawa, Michio; Iwano, Yoshihiko; Ito, Kenichi; Saito, Hiroaki; Kashima, Koichi; Kinoshita, Motoyasu; Okubo, Tadatsune.

    1985-12-01

    FEMAXI-III is a computer code to predict the thermal and mechanical behavior of a light water fuel rod during its irradiation life. It can analyze the integral behavior of a whole fuel rod throughout its life, as well as the localized behavior of a small part of fuel rod. The localized mechanical behavior such as the cladding ridge deformation is analyzed by the two-dimensional axisymmetric finite element method. FEMAXI-III calculates, in particular, the temperature distribution, the radial deformation, the fission gas release, and the inner gas pressure as a function of irradiation time and axial position, and the stresses and strains in the fuel and cladding at a small part of fuel rod as a function of irradiation time. For this purpose, Elasto-plasticity, creep, thermal expansion, fuel cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, and fuel-cladding mechanical interaction are modelled and their interconnected effects are considered in the code. Efforts have been made to improve the accuracy and stability of finite element solution and to minimize the computer memory and running time. This report describes the outline of the code and the basic models involved, and also includes the application of the code and its input manual. (author)

  12. Development of multimedia computer-based training for VXI integrated fuel monitors

    International Nuclear Information System (INIS)

    Keeffe, R.; Ellacott, T.; Truong, Q.S.

    1999-01-01

    The Canadian Safeguards Support Program has developed the VXI Integrated Fuel Monitor (VFIM) which is based on the international VXI instrument bus standard. This equipment is a generic radiation monitor which can be used in an integrated mode where several detection systems can be connected to a common system where information is collected, displayed, and analyzed via a virtual control panel with the aid of computers, trackball and computer monitor. The equipment can also be used in an autonomous mode as a portable radiation monitor with a very low power consumption. The equipment has been described at previous international symposia. Integration of several monitoring systems (bundle counter, core discharge monitor, and yes/no monitor) has been carried out at Wolsong 2. Performance results from one of the monitoring systems which was installed at CANDU nuclear stations are discussed in a companion paper at this symposium. This paper describes the development of an effective multimedia computer-based training package for the primary users of the equipment; namely IAEA inspectors and technicians. (author)

  13. Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Mays, Brian [AREVA Federal Services, Lynchburg, VA (United States); Jackson, R. Brian [TerraPower, Bellevue, WA (United States)

    2017-03-08

    The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services. The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.

  14. Flow field measurements using LDA and numerical computation for rod bundle of reactor fuel assembly

    International Nuclear Information System (INIS)

    Hu Jun; Zou Zunyu

    1995-02-01

    Local mean velocity and turbulence intensity measurements were obtained with DANTEC 55 X two-dimensional Laser Dopper Anemometry (LDA) for rod bundle of reactor fuel assembly test model which was a 4 x 4 rod bundle. The data were obtained from different experimental cross-sections both upstream and downstream of the model support plate. Measurements performed at test Reynolds numbers of 1.8 x 10 4 ∼3.6 x 10 4 . The results described the local and gross effects of the support plate on upstream and downstream flow distributions. A numerical computation was also given, the experimental results are in good agreement with the numerical one and the others in references. Finally, a few suggestions were proposed for how to use the LDA system well. (11 figs.)

  15. Uranium accountability for ATR fuel fabrication: Part II. A computer simulation

    International Nuclear Information System (INIS)

    Dolan, C.A.; Nieschmidt, E.B.; Vegors, S.H. Jr.; Wagner, E.P. Jr.

    1977-08-01

    A stochastic computer model has been designed to simulate the material control system used during the production of fuel plates for the Advanced Test Reactor. Great care has been taken to see that this model follows the manufacturing and measuring processes used. The model is designed so that manufacturing process and measurement parameters are fed in as input; hence, changes in the manufacturing process and measurement procedures are easily simulated. Individual operations in the plant are described by program subroutines. By varying the calling sequence of these subroutines, variations in the manufacturing process may be simulated. By using this model values for MUF and LEMUF may be calculated for predetermined plant operating conditions. Furthermore the effect on MUF and LEMUF produced by changing plant operating procedures and measurement techniques may also be examined. A sample calculation simulating one inventory period of the plant's operation is included

  16. Determination of the vibration characteristics of nuclear fuel rods in a fluid flow using multiphysics computation

    International Nuclear Information System (INIS)

    Sbragio, Ricardo

    1999-01-01

    The determination of natural frequencies and displacement Power Spectrum Density (PSD) of fuel rods in a fluid using Computational Fluid Dynamics and Finite Element Methods is presented. The rods are modeled as slender beams subjected to small displacements in a fluid using three-dimensional mesh. The incompressible Navier-Stokes and linear momentum balance equations are solved simultaneously using Spectrum code. Two examples from literature are analyzed. The first consists in one rod in a fluid. The excitation is an impulse force at the rod central node. The second example is a two rod system in a fluid. In this case, the excitation force is random and is determined from a PSD. (author)

  17. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

  18. Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

    An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

  19. Development of a Computer Program for the Integrated Control of the Fuel Homogeneity Measurement System

    International Nuclear Information System (INIS)

    Shin, H. S.; Jang, J. W.; Lee, Y. H.; Oh, S. J.; Park, H. D.; Kim, C. K.

    2005-11-01

    The computer program is developed based on Visual C++, which is equipped with a user-friendly interface of the input/output(I/O) and a display function for the measuring conditions. This program consists of three parts which are the port communication, PLC(Programmable Logic Controller) and the MCA(Multi Channel Analyzer) control parts. The communication type between the CPU of the PLC module box and the computer is selected as be the Rs-232 asynchronous type and the thread method is adapted in the development of the first part of the program. The PLC-related program has been developed so that the data communication between the PLC CPU and the computer could be harmonized with the unique commands which have already been defined in the PLC. The measuring space and time intervals, the start and end ROI(region of interest) values, and the allowable error limitation are input at each measurement in this program. Finally the controlling MCA program has been developed by using Canberra's programming library which contains several files including the head files in which the variable and the function of C++ are declared according to the MCA function. The performance test has been carried out through an application of the developed computer program to the homogeneity measurement system. The gamma counts at 28 measuring points of a fuel rod of 700 mm in length are measured for 50 sec at each point. It was revealed that the measurement results are better than the previous ones in respects of the measurement accuracy and a measurement time saving could be achieved. It was concluded that the gamma measurement system can be improved through equipping it with the developed control program

  20. Development of a computational model applied to a unitary 144 CM2 proton exchange membrane fuel cell

    International Nuclear Information System (INIS)

    Robalinho, Eric

    2009-01-01

    This work presents the development of a numerical computer model and methodology to study and design polymeric exchange membrane fuel cell - PEM. For the validation of experimental results, a sequence of routines, appropriate to fit the data obtained in the laboratory, was described. At the computational implementation it was created a new strategy of coupling two 3-dimensional models to satisfy the requirements of the comprehensive model of the fuel cell, including its various geometries and materials, as well as the various physical and chemical processes simulated. To effective assessment of the real cell analogy with numerical model, numerical studies were carried out. Comparisons with values obtained in the literature, characterization of variables through laboratory experiments and estimates from models already tested in the literature were also performed. Regarding the experimental part, a prototype of a fuel cell unit of 144 cm 2 of geometric area was designed, produced and operated at laboratory with the purpose of validating the numerical computer model proposed, with positive results. The results of simulations for the 2D and 3D geometries proposed are presented in the form of polarization curves, highlighting the catalytic layer model based on the geometry of agglomerates. Parametric and sensitivity studies are presented to illustrate the change in performance of the fuel cell studied. The final model is robust and useful as a tool for design and optimization of PEM type fuel cells in a wide range of operating conditions. (author)

  1. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza

    2014-01-01

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  2. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

  3. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  4. Emission computer tomography on a Dodewaard mixed oxide fuel pin. Comparative PIE work with non-destructive and destructive techniques

    Energy Technology Data Exchange (ETDEWEB)

    Buurveld, H.A.; Dassel, G.

    1993-12-01

    A nondestructive technique as well as a destructive PIE technique have been used to verify the results obtained with a newly 8-e computer tomography (GECT) system. Multi isotope Scanning (MIS), electron probe micro analysis (EPMA) and GECT were used on a mixed oxide (MOX) fuel rod from the Dodewaard reactor with an average burnup of 24 MWd/kg fuel. GECT shows migration of Cs to the periphery of fuel pellets and to radial cracks and pores in the fuel, whereas MIS shows Cs migration to pellet interfaces. The EPMA technique appeared not to be useful to show migration of Cs but, it shows the distribution of fission products from Pu. EPMA clearly shows the distribution of fission products from Pu, but did not reveal the Cs-migration. (orig./HP)

  5. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

  6. A Computer Simulation to Assess the Nuclear Material Accountancy System of a MOX Fuel Fabrication Facility

    International Nuclear Information System (INIS)

    Portaix, C.G.; Binner, R.; John, H.

    2015-01-01

    SimMOX is a computer programme that simulates container histories as they pass through a MOX facility. It performs two parallel calculations: · the first quantifies the actual movements of material that might be expected to occur, given certain assumptions about, for instance, the accumulation of material and waste, and of their subsequent treatment; · the second quantifies the same movements on the basis of the operator's perception of the quantities involved; that is, they are based on assumptions about quantities contained in the containers. Separate skeletal Excel computer programmes are provided, which can be configured to generate further accountancy results based on these two parallel calculations. SimMOX is flexible in that it makes few assumptions about the order and operational performance of individual activities that might take place at each stage of the process. It is able to do this because its focus is on material flows, and not on the performance of individual processes. Similarly there are no pre-conceptions about the different types of containers that might be involved. At the macroscopic level, the simulation takes steady operation as its base case, i.e., the same quantity of material is deemed to enter and leave the simulated area, over any given period. Transient situations can then be superimposed onto this base scene, by simulating them as operational incidents. A general facility has been incorporated into SimMOX to enable the user to create an ''act of a play'' based on a number of operational incidents that have been built into the programme. By doing this a simulation can be constructed that predicts the way the facility would respond to any number of transient activities. This computer programme can help assess the nuclear material accountancy system of a MOX fuel fabrication facility; for instance the implications of applying NRTA (near real time accountancy). (author)

  7. Computation fluid dynamic modelling of natural convection heat flow in unpumped molten salt fuel tubes

    International Nuclear Information System (INIS)

    Leefe, S.; Jackson-Laver, P.; Scott, I.R.

    2015-01-01

    Use of static molten salt nuclear fuel in simple tubes was discarded in 1949 without considering how convection could affect its utility. This poster describes CFD studies showing that such tubes are practical as fuel elements in essentially conventional fuel assemblies. They can achieve power densities above 250kW per liter of fuel salt (higher than PWR's) and do so without causing the tube wall to heat to dangerous levels. This discovery enables the achievement of the many benefits of molten salt fuel while utilizing the highly developed technology, regulatory, non proliferation and safety benefits of current fuel assembly technology. (author)

  8. Experimental and computation method for determination of burnup and isotopic composition of the WWER-440 fuel using the 134Cs and 137Cs concentrations

    International Nuclear Information System (INIS)

    Babichev, B.A.; Kozharin, V.V.

    1990-01-01

    An experimental and computational method for determination of burnup and actinoid concentrations in WWER fuel elements using 134 Cs and 137 Cs concentrations in fuel is considered. It is shown that the error in calculation of fuel burnup and U and Pu isotope concentrations in WWER-440 fuel elements is 1.3-4.9% provided that the error in 134 Cs and 137 Cs concentration measurements does not exceed 1.7 and 1.2%. 9 refs.; 10 figs.; 4 tabs

  9. Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Regis; Silva, Antonio Teixeira e, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The objective of this paper is to verify the validity and accuracy of the results provided by computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods, in steady-state and transient operational conditions at high burnup. To achieve this goal, the results provided by these computer simulations are compared with experimental data available in the database FUMEX III. Through the results, it was found that the computer programs used have a good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under the influence of Reactivity Initiated Accident (RIA). (author)

  10. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000 degree F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion (''bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled

  11. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  12. A COMPUTATIONAL FLUID DYNAMICS ANALYSIS OF AIR FLOW THROUGH A TELECOM BACK-UP UNIT POWERED BY AN AIR-COOLED PROTON EXCHANGE MEMBRANE FUEL CELL

    DEFF Research Database (Denmark)

    Gao, Xin; Berning, Torsten; Kær, Søren Knudsen

    2016-01-01

    Proton exchange membrane fuel cells (PEMFC’s) are currently being commercialized for various applications ranging from automotive to stationary such as powering telecom back-up units. In PEMFC’s, oxygen from air is internally combined with hydrogen to form water and produce electricity and heat....... This product heat has to be effectively removed from the fuel cell, and while automotive fuel cells are usually liquid-cooled using a secondary coolant loop similar to the internal combustion engines, stationary fuel cell systems as they are used for telecom back-up applications often rely on excessive air fed...... to the fuel cell cathode to remove the heat. Thereby, the fuel cell system is much simpler and cheaper while the fuel cell performance is substantially lower compared to automotive fuel cells. This work presents a computational fluid dynamics analysis on the heat management of an air-cooled fuel cell powered...

  13. Computational and experimental analysis of causes for local deformation of research reactor U-Mo fuel pin claddings in case of high burn-ups

    International Nuclear Information System (INIS)

    Popov, V.V.; Khmelevsky, M.Ya.; Lukichev, V.A.; Golosov, O.A.

    2005-01-01

    Post-reactor investigations of (U-Mo) fuel pins irradiated in the IVV-2M reactor have allowed to determine: the change in a fuel pin volume; the dimensions and the kind of the local deformation of fuel pin claddings; the amount of gases released under the cladding from the fuel composition, the thickness and appearance of the interaction layer of between the (U-Mo) particles and aluminium as a matrix material. The computational analysis of the stressed-strained state of fuel pins has shown that the major contribution to the increase of the fuel pin volume is made by the fuel swelling caused by the solid products of fission being formed in the process of operation. The emergence of the (U-Mo) fuel-aluminium matrix interaction layers around the (U-Mo) particles results in formation and evolution of lamination cavities inside the fuel composition under the joint action of the pressure of process gases and gaseous fission products. In case of high burn-up a local bulge of a fuel pin cladding is being formed in the fuel lamination area caused by the pressure of gases in the presence of creep in the fuel pin cladding material. The computational results relating to the local strain in a research reactor (U-Mo) fuel pin are in a good accordance with the results of the post-reactor investigations. (author)

  14. Computer simulation of the behaviour and performance of a CANDU fuel rod

    International Nuclear Information System (INIS)

    Marino, A.C.

    1997-01-01

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  15. PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

    International Nuclear Information System (INIS)

    Anand, A.K.; Anantharaman, K.; Sarda, V.

    1978-01-01

    The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

  16. A computer aided solvent extraction process design in nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Britto, S.E.; Purandare, H.D.; Lawande, S.V.

    1977-01-01

    A rigorous and conceptual design is attempted of the first step of flowsheet formulation for solvent extraction process for fuel reprocessing plant. The design incorporates three cycles of extraction contractors; the first optimised to maximise Pu recovery while the second and third cycles to maximise fission product decontaminations. There are three basic types of extraction steps in these different cycles requiring painstaking design, namely, extraction-scrub, Pu strip-scrub and simple strip. The extraction system to start with is: U nitrate - Pu nitrate - fission product nitrates - nitric acid - tri-butyl phosphate/diluent. With suitable simplifying assumptions and adopting the concept of discrete equilibrium stagewise operation, simple X-Y operating diagrams could be used. The calculations could therefore be done using McCabe Thiele graphical method. The procedure adopted was to consider the macro-component of U to obtain initial optimum flow sheet details and the number of theoretical stages for each contactor and later to incorporate the behaviour of Pu and fission products. A computer program was written to calculate, for different combinations of nitric acid salting strengths, (1) the U concentration profiles along the contractors and (2) the number of stages needed for various different solvent and aqueous phase flow ratios, using experimentally obtained equilibrium data. The method used is indicated and some samples of results obtained for three types of extraction-scrub operation studied are given. These simplified calculations provided the necessary insight into these difficult operations. (auth.)

  17. Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

    International Nuclear Information System (INIS)

    Shekhar Kumar; Koganti, S.B.

    2003-07-01

    Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

  18. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

    International Nuclear Information System (INIS)

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code

  19. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 1, Mathematical models and solution method

    International Nuclear Information System (INIS)

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations: however, the transient capability has not yet been validated. This volume describes the finite-volume equations and the method used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of these methods

  20. Modeling of the WWER-1000 fuel-rod behavior in steady-state condition with FRAPCONE-3 computer code

    International Nuclear Information System (INIS)

    Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan

    2008-01-01

    It is presented within the paper the results of the modeling and the assessment of the integral code predictions of the WWER fuel-rod behavior in steady-state condition. The assessments in this paper have used the MASSIH and ANS 5.4 subroutine in the code. The modeling and calculations have been performed with FRAPCONE-3 computer code in Argonne National Laboratory, USA

  1. Computer model for refinery operations with emphasis on jet fuel production. Volume 3: Detailed systems and programming documentation

    Science.gov (United States)

    Dunbar, D. N.; Tunnah, B. G.

    1978-01-01

    The FORTRAN computing program predicts flow streams and material, energy, and economic balances of a typical petroleum refinery, with particular emphasis on production of aviation turbine fuels of varying end point and hydrogen content specifications. The program has a provision for shale oil and coal oil in addition to petroleum crudes. A case study feature permits dependent cases to be run for parametric or optimization studies by input of only the variables which are changed from the base case.

  2. 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

  3. SAFE: A computer code for the steady-state and transient thermal analysis of LMR fuel elements

    International Nuclear Information System (INIS)

    Hayes, S.L.

    1993-12-01

    SAFE is a computer code developed for both the steady-state and transient thermal analysis of single LMR fuel elements. The code employs a two-dimensional control-volume based finite difference methodology with fully implicit time marching to calculate the temperatures throughout a fuel element and its associated coolant channel for both the steady-state and transient events. The code makes no structural calculations or predictions whatsoever. It does, however, accept as input structural parameters within the fuel such as the distributions of porosity and fuel composition, as well as heat generation, to allow a thermal analysis to be performed on a user-specified fuel structure. The code was developed with ease of use in mind. An interactive input file generator and material property correlations internal to the code are available to expedite analyses using SAFE. This report serves as a complete design description of the code as well as a user's manual. A sample calculation made with SAFE is included to highlight some of the code's features. Complete input and output files for the sample problem are provided

  4. Comprehensive Analysis of Trends and Emerging Technologies in All Types of Fuel Cells Based on a Computational Method

    Directory of Open Access Journals (Sweden)

    Takaya Ogawa

    2018-02-01

    Full Text Available Fuel cells have been attracting significant attention recently as highly efficient and eco-friendly energy generators. Here, we have comprehensively reviewed all types of fuel cells using computational analysis based on a citation network that detects emerging technologies objectively and provides interdisciplinary data to compare trends. This comparison shows that the technologies of solid oxide fuel cells (SOFCs and electrolytes in polymer electrolyte fuel cells (PEFCs are at the mature stage, whereas those of biofuel cells (BFCs and catalysts in PEFCs are currently garnering attention. It does not mean, however, that the challenges of SOFCs and PEFC electrolytes have been overcome. SOFCs need to be operated at lower temperatures, approximately 500 °C. Electrolytes in PEFCs still suffer from a severe decrease in proton conductivity at low relative humidity and from their high cost. Catalysts in PEFCs are becoming attractive as means to reduce the platinum catalyst cost. The emerging technologies in PEFC catalysts are mainly heteroatom-doped graphene/carbon nanotubes for metal-free catalysts and supports for iron- or cobalt-based catalysts. BFCs have also received attention for wastewater treatment and as miniaturized energy sources. Of particular interest in BFCs are membrane reactors in microbial fuel cells and membrane-less enzymatic biofuel cells.

  5. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.

    2014-01-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10"-"4 to 10"-"5) of as manufactured defects and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from intentionally failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application was considered. Previous experience utilizing similar techniques, the expected activities in AGR-3/4 rings, and analysis of this work indicate using GECT to evaluate AGR-3/4 will be feasible. The GECT technique was also applied to other irradiated nuclear fuel systems currently available in the HFEF hot cell, including oxide fuel pins, metallic fuel pins, and monolithic plate fuel. Results indicate GECT with the HFEF PGS is effective. (author)

  6. CASKETSS-2: a computer code system for thermal and structural analysis of nuclear fuel shipping casks (version 2)

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1991-08-01

    A computer program CASKETSS-2 has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS-2 means a modular code system for CASK Evaluation code system Thermal and Structural Safety (Version 2). Main features of CASKETSS-2 are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) There are simplified computer programs and a detailed one in the structural analysis part in the code system. (3) Input data generator is provided in the code system. (4) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  7. Computer program for modelling the history of the in-service bending of fast power reactor fuel assemblies

    International Nuclear Information System (INIS)

    Dienstbier, J.

    1979-04-01

    The studies into stresses and deformations in the core are mainly focused on the fuel rod and the fuel assembly can. In high neutron doses austenitic steel swells and this is associated with a considerable increase in the volume of material. The SANDRA computer program is used for solving the problems of can deformations and stress during long-term reactor operation. The block for the mechanical interaction of cans is the key part of the program. The program input data include temperature distribution, fast neutron flux distribution and coolant overpressure inside the cans. Reactor operation is modelled using operating modes A, B, C which may arbitrarily be combined. Mode A computes bending deformations and the deformations of the can cross-section due to temperature dilatation in the change in temperature fields in the reactor; mode B computes deformations due to swelling and creep in long-term operation; mode C computes thermal deformations in reactor shut-down. A flowsheet is shown of program SANDRA as are examples of computed deformations. (M.S.)

  8. Computational analysis of modern HTGR fuel performance and fission product release during the HFR-EU1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl, E-mail: k.verfondern@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Xhonneux, André, E-mail: xhonneux@lrst.rwth-aachen.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Nabielek, Heinz, E-mail: heinznabielek@me.com [Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Allelein, Hans-Josef, E-mail: h.j.allelein@fz-juelich.de [Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, Chair for Reactor Safety and Reactor Technology, 52072 Aachen (Germany)

    2014-07-01

    Highlights: • HFR-EU1 irradiation test demonstrates high quality of HTGR spherical fuel elements. • Irradiation performance is in good agreement with German fuel performance modeling. • International benchmark exercise expected first particle to fail at ∼13–17% FIMA. • EOL silver release is predicted to be in the percentage range. • EOL cesium and strontium are expected to remain at a low level. - Abstract: Various countries engaged in the development and fabrication of modern HTGR fuel have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under HTGR operating and accident conditions. Verification and validation studies are conducted by code-to-code benchmarking and code-to-experiment comparisons as part of international exercises. The methodology developed in Germany since the 1980s represents valuable and efficient tools to describe fission product release from spherical fuel elements and TRISO fuel performance, respectively, under given conditions. Continued application to new results of irradiation and accident simulation testing demonstrates the appropriateness of the models in terms of a conservative estimation of the source term as part of interactions with HTGR licensing authorities. Within the European irradiation testing program for HTGR fuel and as part of the former EU RAPHAEL project, the HFR-EU1 irradiation experiment explores the potential for high performance of the presently existing German and newly produced Chinese fuel spheres under defined conditions up to high burnups. The fuel irradiation was completed in 2010. Test samples are prepared for further postirradiation examinations (PIE) including heatup simulation testing in the KÜFA-II furnace at the JRC-ITU, Karlsruhe, to be conducted within the on-going ARCHER Project of the European Commission. The paper will describe the application of the German computer models to the HFR-EU1 irradiation test and

  9. Guide to improving the performance of a manipulator system for nuclear fuel handling through computer controls. Final report

    International Nuclear Information System (INIS)

    Evans, J.M. Jr.; Albus, J.S.; Barbera, A.J.; Rosenthal, R.; Truitt, W.B.

    1975-11-01

    The Office of Developmental Automation and Control Technology of the Institute for Computer Sciences and Technology of the National Bureau of Standards provides advising services, standards and guidelines on interface and computer control systems, and performance specifications for the procurement and use of computer controlled manipulators and other computer based automation systems. These outputs help other agencies and industry apply this technology to increase productivity and improve work quality by removing men from hazardous environments. In FY 74 personnel from the Oak Ridge National Laboratory visited NBS to discuss the feasibility of using computer control techniques to improve the operation of remote control manipulators in nuclear fuel reprocessing. Subsequent discussions led to an agreement for NBS to develop a conceptual design for such a computer control system for the PaR Model 3000 manipulator in the Thorium Uranium Recycle Facility (TURF) at ORNL. This report provides the required analysis and conceptual design. Complete computer programs are included for testing of computer interfaces and for actual robot control in both point-to-point and continuous path modes

  10. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  11. PATE - a computer code for the calculation of temperature distribution in cylindrical fuel rods

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Roberty, N.C.; Carmo, E.G.D. do.

    1983-08-01

    An analytical solution for the temperature profile in the fuel cladding is presented, having the coolant temperature as boundary conditions and using a first-order polynomial for the zircalloy thermal conductivity. The temperature profile in the fuel pellet is determined solving an algebraic equation by iterative methods. (E.G.) [pt

  12. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    Science.gov (United States)

    Poulson, D.; Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J. D.; Plaud-Ramos, K.; Morley, D.; Hecht, A. A.

    2017-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ∼ 18 σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Potential detector technologies and geometries are discussed.

  13. Computer code SICHTA-85/MOD 1 for thermohydraulic and mechanical modelling of WWER fuel channel behaviour during LOCA and comparison with original version of the SICHTA code

    International Nuclear Information System (INIS)

    Bujan, A.; Adamik, V.; Misak, J.

    1986-01-01

    A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)

  14. 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.

  15. SATURN-FS 1: A computer code for thermo-mechanical fuel rod analysis

    International Nuclear Information System (INIS)

    Ritzhaupt-Kleissl, H.J.; Heck, M.

    1993-09-01

    The SATURN-FS code was written as a general revision of the SATURN-2 code. SATURN-FS is capable to perform a complete thermomechanical analysis of a fuel pin, with all thermal, mechanical and irradiation-based effects. Analysis is possible for LWR and for LMFBR fuel pins. The thermal analysis consists of calculations of the temperature profile in fuel, gap and in the cladding. Pore migration, stoichiometry change of oxide fuel, gas release and diffusion effects are taken into account. The mechanical modeling allows the non steady-state analysis of elastic and nonelastic fuel pin behaviour, such as creep, strain hardening, recovery and stress relaxation. Fuel cracking and healing is taken into account as well as contact and friction between fuel and cladding. The modeling of the irradiation effects comprises swelling and fission gas production, Pu-migration and irradiation induced creep. The code structure, the models and the requirements for running the code are described in the report. Recommendations for the application are given. Program runs for verification and typical examples of application are given in the last part of this report. (orig.) [de

  16. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

    International Nuclear Information System (INIS)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-01-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  17. FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup

    International Nuclear Information System (INIS)

    Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

    1997-12-01

    FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes' integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6)

  18. Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

    International Nuclear Information System (INIS)

    Valach, M.; Zymak, J.; Svoboda, R.

    1997-01-01

    This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs

  19. Computer modelling of the WWER fuel elements under high burnup conditions by the computer codes PIN-W and RODQ2D

    Energy Technology Data Exchange (ETDEWEB)

    Valach, M; Zymak, J; Svoboda, R [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

    1997-08-01

    This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs.

  20. MAT-FLX: a simplified code for computing material balances in fuel cycle

    International Nuclear Information System (INIS)

    Pierantoni, F.; Piacentini, F.

    1983-01-01

    This work illustrates a calculation code designed to provide a materials balance for the electro nuclear fuel cycle. The calculation method is simplified but relatively precise and employs a progressive tabulated data approach

  1. Feasibility studies of computed tomography in partial defect detection of spent BWR fuel

    International Nuclear Information System (INIS)

    Levai, F.; Tikkinen, J.; Tarvainen, M.; Arlt, R.

    1990-10-01

    Feasibility studies were made for tomographic reconstruction of a cross-sectional activity distribution of a spent nuclear fuel assembly. The purpose was to determine the number of fuel rods (pins) and localize the positisons where pins are missing. The activity distribution map showing the locations of fuel rods in the assembly was reconstructed. The theoretical part of this work consists of simulation of image reconstruction based on theoretically calculated data from a reference assembly model. Evaluation of different image reconstruction techniques was made. Measurements were made in real facility conditions. Gamma radiation from an irradiated 8 x 8 - 1 BWR fuel assembly was measured through a narrow custom made collimator from different angles and positions. The measured data set was used as projections for reconstructing the activity profile of the assembly in cross-sectional plane

  2. A Computational Study of A Lithium Deuteride Fueled Electrothermal Plasma Mass Accelerator

    OpenAIRE

    Gebhart III, Gerald Edward

    2013-01-01

    Future magnetic fusion reactors such as tokamaks will need innovative, fast, deep-fueling systems to inject frozen deuterium-tritium pellets at high speeds and high repetition rates into the hot plasma core. There have been several studies and concepts for pellet injectors generated, and different devices have been proposed. In addition to fueling, recent studies show that it may be possible to disrupt edge localized mode (ELM) formation by injecting pellets or gas into the fusion plasma. The...

  3. Effects of pressure and fuel dilution on coflow laminar methane-air diffusion flames: A computational and experimental study

    Science.gov (United States)

    Cao, Su; Ma, Bin; Giassi, Davide; Bennett, Beth Anne V.; Long, Marshall B.; Smooke, Mitchell D.

    2018-03-01

    In this study, the influence of pressure and fuel dilution on the structure and geometry of coflow laminar methane-air diffusion flames is examined. A series of methane-fuelled, nitrogen-diluted flames has been investigated both computationally and experimentally, with pressure ranging from 1.0 to 2.7 atm and CH4 mole fraction ranging from 0.50 to 0.65. Computationally, the MC-Smooth vorticity-velocity formulation was employed to describe the reactive gaseous mixture, and soot evolution was modelled by sectional aerosol equations. The governing equations and boundary conditions were discretised on a two-dimensional computational domain by finite differences, and the resulting set of fully coupled, strongly nonlinear equations was solved simultaneously at all points using a damped, modified Newton's method. Experimentally, chemiluminescence measurements of CH* were taken to determine its relative concentration profile and the structure of the flame front. A thin-filament ratio pyrometry method using a colour digital camera was employed to determine the temperature profiles of the non-sooty, atmospheric pressure flames, while soot volume fraction was quantified, after evaluation of soot temperature, through an absolute light calibration using a thermocouple. For a broad spectrum of flames in atmospheric and elevated pressures, the computed and measured flame quantities were examined to characterise the influence of pressure and fuel dilution, and the major conclusions were as follows: (1) maximum temperature increases with increasing pressure or CH4 concentration; (2) lift-off height decreases significantly with increasing pressure, modified flame length is roughly independent of pressure, and flame radius decreases with pressure approximately as P-1/2; and (3) pressure and fuel stream dilution significantly affect the spatial distribution and the peak value of the soot volume fraction.

  4. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    International Nuclear Information System (INIS)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-01

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  5. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  6. Calculation of DND-signals in case of fuel pin failures in KNK II with the computer code FICTION III

    International Nuclear Information System (INIS)

    Schmuck, I.

    1990-11-01

    In KNK II two delayed neutron detectors are installed for quick detection of fuel subassembly cladding failures. They record the release of the precursors of the emitters of delayed neutrons into the sodium. The computer code FICTION III calculates the expected delayed neutron signals for certain fuel pin failures, where the user has to set the boundary conditions interactively. In view of FICTION II the advancement of FICTION III consists of the following items: application of the data sets of 105 isotopes, distinction of thermal and fast neutron induced fission, partitioning of the sodium flow into two circuits, consideration of the specific fission rates in 10 fuel pin sections, elaboration of the user's interaction possibilities for input/ output. The capability of FICTION III is shown by means of two applications (UNi-test pin on position 100 and the third KNK fuel subassembly cladding failure). Object of further evaluations will be among other things the analysis of increased delayed neutron signals in regard to the fault location and dimension

  7. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

    International Nuclear Information System (INIS)

    Tentner, A.

    2009-01-01

    A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

  8. A computer code to estimate accidental fire and radioactive airborne releases in nuclear fuel cycle facilities: User's manual for FIRIN

    International Nuclear Information System (INIS)

    Chan, M.K.; Ballinger, M.Y.; Owczarski, P.C.

    1989-02-01

    This manual describes the technical bases and use of the computer code FIRIN. This code was developed to estimate the source term release of smoke and radioactive particles from potential fires in nuclear fuel cycle facilities. FIRIN is a product of a broader study, Fuel Cycle Accident Analysis, which Pacific Northwest Laboratory conducted for the US Nuclear Regulatory Commission. The technical bases of FIRIN consist of a nonradioactive fire source term model, compartment effects modeling, and radioactive source term models. These three elements interact with each other in the code affecting the course of the fire. This report also serves as a complete FIRIN user's manual. Included are the FIRIN code description with methods/algorithms of calculation and subroutines, code operating instructions with input requirements, and output descriptions. 40 refs., 5 figs., 31 tabs

  9. No one can play snooker that fast. SCK•CEN performs computer simulations of nuclear fuel behaviour

    International Nuclear Information System (INIS)

    2014-01-01

    The article discusses the project F-Bridge, which is funded under the seventh European framework programme. The objective of F-Bridge apply knowledge gained from fundamental science to to the development of Gen-IV fuels. The ultimate goal is to set up an international exchange platform for the integration and the coordination of nuclear fuel research. As part of the F-Bridge project, the Belgian Nuclear Research Center SCK-CEN collaborates with other institutes in Europe in a work package on multi-scale modelling. The main contribution of SCK-CEN focuses on atomic-scale computer simulations. The employed techniques are based on empirical force fields and describe atomic interactions using simple parameters and analytical formulas. Based on Newton’s theory, the evolution of a series of atoms and their interaction over a given time period can be predicted.

  10. Computer simulation of thermal-hydraulics of MNSR fuel-channel assembly using LabView

    International Nuclear Information System (INIS)

    Gadri, L. A.

    2013-07-01

    A LabView simulator of thermal hydraulics has been developed to demonstrate the temperature profile of coolant flow in the reactor core during normal operation. The simulator could equally be used for any transient behaviour of the reactor. Heat generation, transfer and the associated temperature profile in the fuel-channel elements viz: the coolant, cladding and fuel were studied and the corresponding analytical temperature equations in the axial and radial directions for the coolant, outer surface of the cladding, fuel surface and fuel center were obtained for the simulation using LabView. Tables of values for the equations were constructed by MATLAB and excel software programs. Plots of the equations with LabView were verified and validated with the graphs drawn by the MATLAB. In this thesis, an analysis of the effects of the coolant inlet temperature of 24.5°C and exit temperature of 70.0° on the temperature distribution in fuel-channel elements of the reactor core of cylindrical geometry was carried out. Other parameters, including the total fuel channel power, mass flow rate and convective heat transfer coefficient were varied to study the effects on the temperature profile. The analytical temperature equations in the fuel channel elements of the reactor core were obtained. MATLAB and Excel software were used to construct data for the equations. The plots by MATLAB were used to benchmark the LabVIEW simulation. Excellent agreement was obtained between the MATLAB plots and the LabView simulation results with an error margin of 0.001. The analysis of the results by comparing gradients of inlet temperature, total reactor channel power and mass flow indicated that inlet temperature gradient is one of the key parameters in determining the temperature profile in the MNSR core. (au)

  11. SCALE-4 [Standardized Computer Analyses for Licensing Evaluation]: An improved computational system for spent-fuel cask analysis

    International Nuclear Information System (INIS)

    Parks, C.V.

    1989-01-01

    The purpose of this paper is to provide specific information regarding improvements available with Version 4.0 of the SCALE system and discuss the future of SCALE within the current computing and regulatory environment. The emphasis focuses on the improvements in SCALE-4 over that available in SCALE-3. 10 refs., 1 fig., 1 tab

  12. ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

    International Nuclear Information System (INIS)

    Kovscek, S.E.; Martin, S.E.

    1982-10-01

    ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given

  13. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

  14. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; Biwer, B.M.; LePoire, D.J.

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows trademark environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident

  15. A computer program for calculation of the fuel cycle in pressurized water reactors

    International Nuclear Information System (INIS)

    Solanilla, R.

    1976-01-01

    The purpose of the FUCEFURE program is two-fold: first, it is designed to solve the problem of nuclear fuel cycle cost in one pressurized light water reactor calculation. The code was developed primarily for comparative and sensitivity studies. The program contains simple correlations between exposure and available depletion data used to predict the uranium and plutonium content of the fuel as a function of the fuel initial enrichment. Second, it has been devised to evaluate the nuclear fuel demand associated with an expanding nuclear power system. Evaluation can be carried out at any time and stage in the fuel cycle. The program can calculate the natural uranium and separate work requirements of any final and tails enrichment. It also can determine the nuclear power share of each reactor in the system when a decision has been made about the long-term nuclear power installations to be used and the types of PWR and fast breeder reactor characteristics to be involved in them. (author)

  16. V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

    International Nuclear Information System (INIS)

    Teuchert, E.; Hansen, U.; Haas, K.A.

    1980-03-01

    V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

  17. User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

    1979-10-01

    The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

  18. Computed isotopic inventory and dose assessment for SRS fuel and target assemblies

    International Nuclear Information System (INIS)

    Chandler, M.C.; Ketusky, E.T.; Thoman, D.C.

    1995-01-01

    Past studies have identified and evaluated important radionuclide contributors to dose from reprocessed spent fuel sent to waste for Mark 16B and 22 fuel assemblies and for Mark 31 A and 31B target assemblies. Fission-product distributions after a 5- and 15-year decay time were calculated for a ''representative'' set of irradiation conditions (i.e., reactor power, irradiation time, and exposure) for each type of assembly. The numerical calculations were performed using the SHIELD/GLASS system of codes. The sludge and supernate source terms for dose were studied separately with the significant radionuclide contributors for each identified and evaluated. Dose analysis considered both inhalation and ingestion pathways: The inhalation pathway was analyzed for both evaporative and volatile releases. Analysis of evaporative releases utilized release fractions for the individual radionuclides as defined in the ICRP-30 by DOE guidance. A release fraction of unity was assumed for each radionuclide under volatile-type releases, which would encompass internally initiated events (e.g., fires, explosions), process-initiated events, and externally initiated events. Radionuclides which contributed at least 1% to the overall dose were designated as significant contributors. The present analysis extends and complements the past analyses through considering a broader spectrum of fuel types and a wider range of irradiation conditions. The results provide for a more thorough understanding of the influences of fuel composition and irradiation parameters on fission product distributions (at 2 years or more). Additionally, the present work allows for a more comprehensive evaluation of radionuclide contributions to dose and an estimation of the variability in the radionuclide composition of the dose source term that results from the spent fuel sent to waste encompassing a broad spectrum of fuel compositions and irradiation conditions

  19. ADVANCING THE FUNDAMENTAL UNDERSTANDING AND SCALE-UP OF TRISO FUEL COATERS VIA ADVANCED MEASUREMENT AND COMPUTATIONAL TECHNIQUES

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, Pratim; Al-Dahhan, Muthanna

    2012-11-01

    to advance the fundamental understanding of the hydrodynamics by systematically investigating the effect of design and operating variables, to evaluate the reported dimensionless groups as scaling factors, and to establish a reliable scale-up methodology for the TRISO fuel particle spouted bed coaters based on hydrodynamic similarity via advanced measurement and computational techniques. An additional objective is to develop an on-line non-invasive measurement technique based on gamma ray densitometry (i.e. Nuclear Gauge Densitometry) that can be installed and used for coater process monitoring to ensure proper performance and operation and to facilitate the developed scale-up methodology. To achieve the objectives set for the project, the work will use optical probes and gamma ray computed tomography (CT) (for the measurements of solids/voidage holdup cross-sectional distribution and radial profiles along the bed height, spouted diameter, and fountain height) and radioactive particle tracking (RPT) (for the measurements of the 3D solids flow field, velocity, turbulent parameters, circulation time, solids lagrangian trajectories, and many other of spouted bed related hydrodynamic parameters). In addition, gas dynamic measurement techniques and pressure transducers will be utilized to complement the obtained information. The measurements obtained by these techniques will be used as benchmark data to evaluate and validate the computational fluid dynamic (CFD) models (two fluid model or discrete particle model) and their closures. The validated CFD models and closures will be used to facilitate the developed methodology for scale-up, design and hydrodynamic similarity. Successful execution of this work and the proposed tasks will advance the fundamental understanding of the coater flow field and quantify it for proper and safe design, scale-up, and performance. Such achievements will overcome the barriers to AGR applications and will help assure that the US maintains

  20. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The code combines a transient 2-dimensional heat conduction code and a 1-dimensional mechanical model for the cladding deformation. The first sections of this report deal with the heat conduction model and the finite element discretization used for the thermal analysis. The mechanical deformation model is presented next: modelling of creep, phase change and oxidation of the zircaloy cladding is discussed in detail. A model describing the effect of oxidation and oxide cracking on the mechanical strength of the cladding is presented too. Next a mechanical restraint model, which allows the simulation of the presence of the neighbouring rods and is particularly important in assessing the amount of channel blockage during a transient, is presented. A description of the models used for the coolant conditions and for the power generation follows. The heat source can be placed either in the fuel or in the cladding, and direct or indirect clad heating by electrical power can be simulated. Then a section follows, dealing with the steady-state and transient types of calculation and with the automatic variable time step selection during the transient. The last sections deal with presentation of results, graphical output, test problems and an example of general application of the code

  1. DATING: A computer code for determining allowable temperatures for dry storage of spent fuel in inert and nitrogen gases

    International Nuclear Information System (INIS)

    Simonen, E.P.; Gilbert, E.R.

    1988-12-01

    The DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) code can be used to calculate allowable initial temperatures for dry storage of light-water-reactor spent fuel. The calculations are based on the life fraction rule using both measured data and mechanistic equations as reported by Chin et al. (1986). The code is written in FORTRAN and utilizes an efficient numerical integration method for rapid calculations on IBM-compatible personal computers. This report documents the technical basis for the DATING calculations, describes the computational method and code statements, and includes a user's guide with examples. The software for the DATING code is available through the National Energy Software Center operated by Argonne National Laboratory, Argonne, Illinois 60439. 5 refs., 8 figs., 5 tabs

  2. CASKETSS-DYNA2D: a nonlinear impact analysis computer program for nuclear fuel transport casks in two dimensional geometries

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-10-01

    A nonlinear impact analysis computer program DYNA2D, which was developed by Hallquist, has been introduced from Lawrence Livermore National Laboratory for the purpose of using impact analysis of nuclear fuel transport casks. DYNA2D has been built in CASKETSS code system (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of DYNA2D are as follows; (1) This program has been programmed to provide near optimal speed on vector processing computers. (2) An explicit time integration method is used for fast calculation. (3) Many material models are available in the program. (4) A contact-impact algorithm permits gap and sliding along structural interfaces. (5) A rezoner has been embedded in the program. (6) The graphic program for representations of calculation is provided. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  3. A computer code for calculation of solvent-extraction separation in a multicomponent system with reference to nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    Carassiti, F.; Liuzzo, G.; Morelli, A.

    1982-01-01

    Nuclear technology development pointed out the need for a new assessment of the fuel cycle back-end. Treatment and disposal of radioactive wastes arising from nuclear fuel reprocessing is known as one of the problems not yet satisfactorily solved, together with separation process of uranium and plutonium from fission products in highly irradiated fuels. Aim of this work is to present an improvement of the computer code for solvent extraction process calculation previously designed by the authors. The modeling of the extraction system has been modified by introducing a new method for calculating the distribution coefficients. The new correlations were based on deriving empirical functions for not only the apparent equilibrium constants, but also the solvation number. The mathematical model derived for calculating separation performance has been then tested for up to ten components and twelve theoretical stages with minor modifications to the convergence criteria. Suitable correlations for the calculation of the distribution coefficients of Uranium, Plutonium, Nitric Acid and fission products were constructed and used to successfully simulate several experimental conditions. (Author)

  4. Computer Simulations of Composite Electrodes in Solid-Oxide Fuel-Cells

    Energy Technology Data Exchange (ETDEWEB)

    Sunde, Svein

    1999-07-01

    Fuel cells are devices for converting the combined chemical (free) energy of fuels and oxygen (air) directly to electrical energy without relying on the dynamic action of steam heated by reacting fuel-oxygen mixtures, like in steam turbines, or of the reacting gas mixtures themselves, like in gas turbines. The basic rationale for fuel cells is their high efficiencies as compared to indirect-conversion methods. Fuel cells are currently being considered for a number of applications, among them de-centralised power supply. Fuel cells come in five basic types and are usually classified according to the type of electrolyte used, which in turn to a significant degree limits the options for anode and cathode materials. The solid-oxide fuel-cell (SOFC) , with which this thesis is concerned, is thus named after its oxide electrolyte, typically the oxide-ion conducting material yttria-stabilised zirconia (YSZ). While the cathode of an SOFC is often uniform in chemical composition (or at least intended to be), various problems of delamination, cracking etc. associated with the use of metallic anode electrocatalysts led to the development of composite SOFC anodes. Porous anodes consisting of Ni and YSZ particles in roughly 50/50 wt-% mixtures are now almost standard with any SOFC-development programme. The designer of composite SOFC electrodes is faced with at least three, interrelated questions: (1) What will be the optimum microstructure and composition of the composite electrode? (2) If the structure changes during operation, as is often observed, what will be the consequences for the internal losses in the cell? (3) How do we interpret electrochemical and conductivity measurements with regard to structure and composition? It is the primary purpose of this thesis to provide a framework for modelling the electrochemical and transport properties of composite electrodes for SOFC, and to arrive at some new insights that cannot be offered by experiment alone. Emphasis is put on

  5. Fuel assemblies mechanical behaviour improvements based on design changes and loading patterns computational analyses

    International Nuclear Information System (INIS)

    Marin, J.; Aullo, M.; Gutierrez, E.

    2001-01-01

    In the past few years, incomplete RCCA insertion events (IRI) have been taking place at some nuclear plants. Large guide thimble distortion caused by high compressive loads together with the irradiation induced material creep and growth, is considered as the primary cause of those events. This disturbing phenomenon is worsened when some fuel assemblies are deformed to the extent that they push the neighbouring fuel assemblies and the distortion is transmitted along the core. In order to better understand this mechanism, ENUSA has developed a methodology based on finite element core simulation to enable assessments on the propensity of a given core loading pattern to propagate the distortion along the core. At the same time, the core loading pattern could be decided interacting with nuclear design to obtain the optimum response under both, nuclear and mechanical point of views, with the objective of progressively attenuating the core distortion. (author)

  6. Evaluation of spent fuel isotopics, radiation spectra and decay heat using the scale computational system

    International Nuclear Information System (INIS)

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

    1986-01-01

    In order to be a self-sufficient system for transport/storage cask shielding and heat transfer analysis, the SCALE system developers included modules to evaluate spent fuel radiation spectra and decay heat. The primary module developed for these analyses is ORIGEN-S which is an updated verision of the original ORIGEN code. The COUPLE module was also developed to enable ORIGEN-S to easily utilize multigroup cross sections and neutron flux data during a depletion analysis. Finally, the SAS2 control module was developed for automating the depletion and decay via ORIGEN-S while using burnup-dependent neutronic data based on a user-specified fuel assembly and reactor history. The ORIGEN-S data libraries available for depletion and decay have also been significantly updated from that developed with the original ORIGEN code

  7. FEMAXI-III, a computer code for fuel rod performance analysis

    International Nuclear Information System (INIS)

    Ito, K.; Iwano, Y.; Ichikawa, M.; Okubo, T.

    1983-01-01

    This paper presents a method of fuel rod thermal-mechanical performance analysis used in the FEMAXI-III code. The code incorporates the models describing thermal-mechanical processes such as pellet-cladding thermal expansion, pellet irradiation swelling, densification, relocation and fission gas release as they affect pellet-cladding gap thermal conductance. The code performs the thermal behavior analysis of a full-length fuel rod within the framework of one-dimensional multi-zone modeling. The mechanical effects including ridge deformation is rigorously analyzed by applying the axisymmetric finite element method. The finite element geometrical model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The 8-node quadratic isoparametric ring elements are adopted for obtaining accurate finite element solutions. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behaviors accurately and stably. The pellet-cladding interaction mechanism is exactly treated using the nodal continuity conditions. The code is applicable to the thermal-mechanical analysis of water reactor fuel rods experiencing variable power histories. (orig.)

  8. Seismic justification of free standing spent fuel storage racks: experimental versus computed behavior

    International Nuclear Information System (INIS)

    Champomier, F.P.; Peron, J.Y.

    2001-01-01

    Most spent fuel storage racks now designed for nuclear power plants are of the free standing type. They consist in a series of independent modules of varying size (typically, 2 m long and wide and 5 m high) immersed in water but simply left standing on the bottom of the spent fuel storage pool. The ever increasing need for additional storage room leads power plants to wish to have compact modules positioned as close to each other as possible. The reduced clearance is essentially limited by the possible interaction of a module with its neighbors in case of a seismic event. In fact, during an earthquake, it is anticipated that each module may have a complex and different movement relative to the pool floor: sliding, uplifting, rocking may occur depending on various parameters related to module geometry and load and seismic excitation. Minimum clearance between modules is to be defined and justified. Two approaches are possible: each module may be demonstrated to stay wide from its neighbors or to impact them with no significant resulting damage. Though experimental work is sometimes possible, it certainly cannot be routinely or systematically envisaged. Thus, approaches to evaluation and justification basically rely on numerical simulation of the problem. Actually, the entire calculation scheme is often repeated a large number of times with varying parameters (seismic input in the three directions, friction coefficients, fuel loading characteristics, etc.) for additional treatment and confidence. The object of this paper is to provide some insight on the kind of information one can reasonably expect to obtain from such studies. (author)

  9. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  10. Computer program for obtaining thermodynamic and transport properties of air and products of combustion of ASTM-A-1 fuel and air

    Science.gov (United States)

    Hippensteele, S. A.; Colladay, R. S.

    1978-01-01

    A computer program for determining desired thermodynamic and transport property values by means of a three-dimensional (pressure, fuel-air ratio, and either enthalpy or temperature) interpolation routine was developed. The program calculates temperature (or enthalpy), molecular weight, viscosity, specific heat at constant pressure, thermal conductivity, isentropic exponent (equal to the specific heat ratio at conditions where gases do not react), Prandtl number, and entropy for air and a combustion gas mixture of ASTM-A-1 fuel and air over fuel-air ratios from zero to stoichiometric, pressures from 1 to 40 atm, and temperatures from 250 to 2800 K.

  11. A computational study of droplet evaporation with fuel vapor jet ejection induced by localized heat sources

    KAUST Repository

    Sim, Jaeheon

    2015-05-12

    Droplet evaporation by a localized heat source under microgravity conditions was numerically investigated in an attempt to understand the mechanism of the fuel vapor jet ejection, which was observed experimentally during the flame spread through a droplet array. An Eulerian-Lagrangian method was implemented with a temperature-dependent surface tension model and a local phase change model in order to effectively capture the interfacial dynamics between liquid droplet and surrounding air. It was found that the surface tension gradient caused by the temperature variation within the droplet creates a thermo-capillary effect, known as the Marangoni effect, creating an internal flow circulation and outer shear flow which drives the fuel vapor into a tail jet. A parametric study demonstrated that the Marangoni effect is indeed significant at realistic droplet combustion conditions, resulting in a higher evaporation constant. A modified Marangoni number was derived in order to represent the surface force characteristics. The results at different pressure conditions indicated that the nonmonotonic response of the evaporation rate to pressure may also be attributed to the Marangoni effect.

  12. Homogenized parameters of light water fuel elements computed by a perturbative (perturbation) method

    International Nuclear Information System (INIS)

    Koide, Maria da Conceicao Michiyo

    2000-01-01

    A new analytic formulation for material parameters homogenization of the two dimensional and two energy-groups diffusion model has been successfully used as a fast computational tool for recovering the detailed group fluxes in full reactor cores. The homogenization method which has been proposed does not require the solution of the diffusion problem by a numerical method. As it is generally recognized that currents at assembly boundaries must be computed accurately, a simple numerical procedure designed to improve the values of currents obtained by nodal calculations is also presented. (author)

  13. Influence of the operational parameters on bioelectricity generation in continuous microbial fuel cell, experimental and computational fluid dynamics modelling

    Science.gov (United States)

    Sobieszuk, Paweł; Zamojska-Jaroszewicz, Anna; Makowski, Łukasz

    2017-12-01

    The influence of the organic loading rate (also known as active anodic chamber volume) on bioelectricity generation in a continuous, two-chamber microbial fuel cell for the treatment of synthetic wastewater, with glucose as the only carbon source, was examined. Ten sets of experiments with different combinations of hydraulic retention times (0.24-1.14 d) and influent chemical oxygen demand concentrations were performed to verify the impact of organic loading rate on the voltage generation capacity of a simple dual-chamber microbial fuel cell working in continuous mode. We found that there is an optimal hydraulic retention time value at which the maximum voltage is generated: 0.41 d. However, there were no similar effects, in terms of voltage generation, when a constant hydraulic retention time with different influent chemical oxygen demand of wastewater was used. The obtained maximal voltage value (600 mV) has also been compared to literature data. Computational fluid dynamics (CFD) was used to calculate the fluid flow and the exit age distribution of fluid elements in the reactor to explain the obtained experimental results and identify the crucial parameters for the design of bioreactors on an industrial scale.

  14. User's manual for the CC3 computer models of the concept for disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    Dougan, K.D.; Wojciechowski, L.C.

    1995-06-01

    Atomic Energy of Canada Limited (AECL) is assessing a concept for disposing of CANDU reactor fuel waste in a vault deep in plutonic rock of the Canadian Shield. A computer program called the Systems Variability Analysis Code (SYVAC) has been developed as an analytical tool for the postclosure (long-term) assessment of the concept, and for environmental assessments of other systems. SYVAC3, the third generation of the code, is an executive program that directs repeated simulation of the disposal system, which is represented by the CC3 (Canadian Concept, generation 3) models comprising a design-specific vault, a site-specific geosphere and a biosphere typical of the Canadian Shield. (author). 23 refs., 7 tabs., 21 figs

  15. Three-dimensional multi-phase flow computational fluid dynamics model for analysis of transport phenomena and thermal stresses in PEM fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Maher, A.R.; Al-Baghdadi, S. [International Technological Univ., London (United Kingdom). Dept. of Mechanical Engineering; Haroun, A.K.; Al-Janabi, S. [Babylon Univ., Babylon (Iraq). Dept. of Mechanical Engineering

    2007-07-01

    Fuel cell technology is expected to play an important role in meeting the growing demand for distributed generation because it can convert the chemical energy of a clean fuel directly into electrical energy. An operating fuel cell has varying local conditions of temperature, humidity, and power generation across the active area of the fuel cell in 3D. This paper presented a model that was developed to improve the basic understanding of the transport phenomena and thermal stresses in PEM fuel cells, and to investigate the behaviour of polymer membrane under hygro and thermal stresses during the cell operation. This comprehensive 3D, multiphase, non-isothermal model accounts for the major transport phenomena in a PEM fuel cell, notably convective and diffusive heat and mass transfer; electrode kinetics; transport and phase change mechanism of water; and potential fields. The model accounts for the liquid water flux inside the gas diffusion layers by viscous and capillary forces and can therefore predict the amount of liquid water inside the gas diffusion layers. This study also investigated the key parameters affecting fuel cell performance including geometry, materials and operating conditions. The model considers the many interacting, complex electrochemical, transport phenomena, thermal stresses and deformation that cannot be studied experimentally. It was concluded that the model can provide a computer-aided tool for the design and optimization of future fuel cells with much higher power density and lower cost. 21 refs., 2 tabs., 14 figs.

  16. Conception and development of a computer-aided design for a spent fuel reprocessing plant

    International Nuclear Information System (INIS)

    Blayrac, M.

    1989-06-01

    A spent fuel reprocessing plant is composed of connected equipments. The aim of this study is the creation of schemes representing the different workshops of the plant and the calculation of linkage characteristics (flux) from a graphic description of functional structures. The program, written in FORTRAN 77, based on mass, flow rate and energy conservation, uses a module library each corresponding to an elementary operation of chemical engineering. Verification is necessary for result quality and accuracy. The important number of parameters and variables used in the program, requires a diagnosis accelerating research of errors for correction. Knowledges used in these last operations are qualitative (knowledge of experts) and quantitative (results of calculations) for the development of an expert system written in D-PROLOG [fr

  17. Fuel behaviour and fission product release under realistic hydrogen conditions comparisons between HEVA 06 test results and Vulcain computations

    International Nuclear Information System (INIS)

    Dumas, J.M.; Lhiaubet, G.

    1989-07-01

    The HEVA 06 test was designed to simulate the conditions existing at the time when fission products are released from irradiated fuel under hydrogen conditions occurring in a PWR core at low pressure. The test conditions were defined from results provided by the core degradation module of the ESCADRE system (1): VULCAIN. This computer code has been recently used to analyse the early core degradation of a 900 MWe PWR in the AF accident sequence (as defined in WASH - 1400, USNRC - 1975). In this scenario, the core would begin to uncover about one day after scram with the system pressure at about 0.4 MPa. The fission product release starts 70 minutes after core dewatering. The F.P. are transferred to the core outlet in an increasingly hydrogen-rich steam atmosphere. The carrier gas is nearly pure hydrogen in the time period 100 - 130 minutes after core uncovering. A large release of F.P. is predicted in the upper part of the core when the steam starvation occurs. At that time, two thirds of the cladding have been oxidised on an average. Before each HEVA test a fuel sample with a burn-up of 36 GWd/tU is reirradiated in order to observe the release of short-lived fission products. A pre-oxidation was primarely conducted in the HEVA 06 test at a temperature of 1300 0 C and controlled to reach a 2/3 cladding oxidation state. Then the steam was progressively replaced by hydrogen and a heat-up rate of 1.5 0 C/s was induced to reach a temperature of 2100 0 C. The fuel was maintained at this temperature for half an hour in hydrogen. The volatile F.P. release kinetics were observed by on-line gamma spectrometry. Pre test calculations of F.P. release kinetics performed with the EMIS module based on the CORSOR models (3) are compared with the test results. Measured releases of cesium and iodine are really lower than those predicted. Axial and radial F.P. distributions in the fuel pellets are available from gamma tomography measurements performed after the test. Tellurium seems

  18. Nuclear fuel cycle risk assessment: survey and computer compilation of risk-related literature. [Once-through Cycle and Plutonium Recycle

    Energy Technology Data Exchange (ETDEWEB)

    Yates, K.R.; Schreiber, A.M.; Rudolph, A.W.

    1982-10-01

    The US Nuclear Regulatory Commission has initiated the Fuel Cycle Risk Assessment Program to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. Both the once-through cycle and plutonium recycle are being considered. A previous report generated by this program defines and describes fuel cycle facilities, or elements, considered in the program. This report, the second from the program, describes the survey and computer compilation of fuel cycle risk-related literature. Sources of available information on the design, safety, and risk associated with the defined set of fuel cycle elements were searched and documents obtained were catalogued and characterized with respect to fuel cycle elements and specific risk/safety information. Both US and foreign surveys were conducted. Battelle's computer-based BASIS information management system was used to facilitate the establishment of the literature compilation. A complete listing of the literature compilation and several useful indexes are included. Future updates of the literature compilation will be published periodically. 760 annotated citations are included.

  19. FEMAXI-III. An axisymmetric finite element computer code for the analysis of fuel rod performance

    International Nuclear Information System (INIS)

    Ichikawa, M.; Nakajima, T.; Okubo, T.; Iwano, Y.; Ito, K.; Kashima, K.; Saito, H.

    1980-01-01

    For the analysis of local deformation of fuel rods, which is closely related to PCI failure in LWR, FEMAXI-III has been developed as an improved version based on the essential models of FEMAXI-II, MIPAC, and FEAST codes. The major features of FEMAXI-III are as follows: Elasto-plasticity, creep, pellet cracking, relocation, densification, hot pressing, swelling, fission gas release, and their interrelated effects are considered. Contact conditions between pellet and cladding are exactly treated, where sliding or sticking is defined by iterations. Special emphasis is placed on creep and pellet cracking. In the former, an implicit algorithm is applied to improve numerical stability. In the latter, the pellet is assumed to be non-tension material. The recovery of pellet stiffness under compression is related to initial relocation. Quadratic isoparametric elements are used. The skyline method is applied to solve linear stiffness equation to reduce required core memories. The basic performance of the code has been proven to be satisfactory. (author)

  20. PELLET: a computer routine for modeling pellet fueling in tokamak plasmas

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Iskra, M.A.; Howe, H.C.; Attenberger, S.E.

    1979-01-01

    Recent experimental results of frozen hydrogenic pellet injection into hot tokamak plasmas and substantial agreement with theoretical predictions have led to a much greater interest in pellets as a means of refueling plasmas. The computer routine PELLET has been developed and used as an aid in assessing pellet ablation models and the effects of pellets on plasma behavior. PELLET provides particle source profiles under various options for the ablation model and can be coupled either to a fluid transport code or to a brief routine which supplies the required input parameters

  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

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  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

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  3. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  4. Computational fluid dynamics analysis of an innovative start-up method of high temperature fuel cells using dynamic 3d model

    Directory of Open Access Journals (Sweden)

    Kupecki Jakub

    2017-03-01

    Full Text Available The article presents a numerical analysis of an innovative method for starting systems based on high temperature fuel cells. The possibility of preheating the fuel cell stacks from the cold state to the nominal working conditions encounters several limitations related to heat transfer and stability of materials. The lack of rapid and safe start-up methods limits the proliferation of MCFCs and SOFCs. For that reason, an innovative method was developed and verified using the numerical analysis presented in the paper. A dynamic 3D model was developed that enables thermo-fluidic investigations and determination of measures for shortening the preheating time of the high temperature fuel cell stacks. The model was implemented in ANSYS Fluent computational fluid dynamic (CFD software and was used for verification of the proposed start-up method. The SOFC was chosen as a reference fuel cell technology for the study. Results obtained from the study are presented and discussed.

  5. Computational comparison of the effect of mixing grids of 'swirler' and 'run-through' types on flow parameters and the behavior of steam phase in WWER fuel assemblies

    International Nuclear Information System (INIS)

    Shcherbakov, S.; Sergeev, V.

    2011-01-01

    The results obtained using the TURBOFLOW computer code are presented for the numerical calculations of space distributions of coolant flow, heating and boiling characteristics in WWER fuel assemblies with regard to the effect of mixing grids of 'Swirler' and 'Run-through' types installed in FA on the above processes. The nature of the effect of these grids on coolant flow was demonstrated to be different. Thus, the relaxation length of cross flows after passing a 'Run-through' grid is five times as compared to a 'Swirler'-type grid, which correlates well with the experimental data. At the same time, accelerations occurring in the flow downstream of a 'Swirler'-type grid are by an order of magnitude greater than those after a 'Run-through' grid. As a result, the efficiency of one-phase coolant mixing is much higher for the grids of 'Run-through' type, while the efficiency of steam removal from fuel surface is much higher for 'Swirler'-type grids. To achieve optimal removal of steam from fuel surface it has been proposed to install into fuel assemblies two 'Swirler'-type grids in tandem at a distance of about 10 cm from each other with flow swirling in opposite directions. 'Run-through' grids would be appropriate for use for mixing in fuel assemblies with a high non-uniformity of fuel-by-fuel power generation. (authors)

  6. A computer-aided diagnostic and troubleshooting system for fuel cell power plants

    International Nuclear Information System (INIS)

    Unkle, C.R.

    1990-10-01

    This Interactive Computer-Aided Troubleshooting System (ICATS) was designed as a troubleshooting aid for the Tokyo Electric Power Company (TEPCO) 11-megawatt dc Module (DCM). ICATS represents an integration of the System Testability and Maintenance Program (STAMP reg-sign) and the Portable Interactive Troubleshooter (POINTER trademark) software packages developed by ARINC Research Corporation. ICATS was designed to aid in the fault isolation of shutdown conditions that may occur in the DCM during on-load operations and start-up, and hold conditions that may occur during start-up. ICATS may also be used to help fault-isolate a return-to-standby condition occurring from on-load operation of the DCM. This report describes the development of ICATS and the ICATS functional design. The unique features of STAMP and POINTER, which allow for diagnostic aids to be designed for systems not yet built and operating, are also described. 10 refs., 21 figs., 6 tabs

  7. A Computational and Experimental Study of Coflow Laminar Methane/Air Diffusion Flames: Effects of Fuel Dilution, Inlet Velocity, and Gravity

    Science.gov (United States)

    Cao, S.; Ma, B.; Bennett, B. A. V.; Giassi, D.; Stocker, D. P.; Takahashi, F.; Long, M. B.; Smooke, M. D.

    2014-01-01

    The influences of fuel dilution, inlet velocity, and gravity on the shape and structure of laminar coflow CH4-air diffusion flames were investigated computationally and experimentally. A series of nitrogen-diluted flames measured in the Structure and Liftoff in Combustion Experiment (SLICE) on board the International Space Station was assessed numerically under microgravity (mu g) and normal gravity (1g) conditions with CH4 mole fraction ranging from 0.4 to 1.0 and average inlet velocity ranging from 23 to 90 cm/s. Computationally, the MC-Smooth vorticity-velocity formulation was employed to describe the reactive gaseous mixture, and soot evolution was modeled by sectional aerosol equations. The governing equations and boundary conditions were discretized on a two-dimensional computational domain by finite differences, and the resulting set of fully coupled, strongly nonlinear equations was solved simultaneously at all points using a damped, modified Newton's method. Experimentally, flame shape and soot temperature were determined by flame emission images recorded by a digital color camera. Very good agreement between computation and measurement was obtained, and the conclusions were as follows. (1) Buoyant and nonbuoyant luminous flame lengths are proportional to the mass flow rate of the fuel mixture; computed and measured nonbuoyant flames are noticeably longer than their 1g counterparts; the effect of fuel dilution on flame shape (i.e., flame length and flame radius) is negligible when the flame shape is normalized by the methane flow rate. (2) Buoyancy-induced reduction of the flame radius through radially inward convection near the flame front is demonstrated. (3) Buoyant and nonbuoyant flame structure is mainly controlled by the fuel mass flow rate, and the effects from fuel dilution and inlet velocity are secondary.

  8. Transient heat transfer analysis up to dryout in 3D fuel rods under unideal conditions through the development of a computer code

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Rodolfo I.; Affonso, Renato R.W.; Moreira, Maria de Lourdes; Sampaio, Paulo A. B. de, E-mail: rodolfoienny@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    In this paper we analyze a conjugated transient heat transfer problem consisting of a nuclear reactor's fuel rod and its intrinsic coolant channel. Our analysis is made possible through a computer code being developed at the Instituto de Engenharia Nuclear (IEN/CNEN). This code is meant to study the temperature behavior in fuel rods which exhibit deviation from their ideal conditions, that is, rods in which the cladding is deformed or the fuel is dislocated. It is also designed to avoid the use of the computationally expensive Navier-Stokes equations. For these reasons, its physical model has as basis a three-dimensional fuel rod coupled to a one-dimensional coolant channel, which are discretized using the finite element method. Intending to study accidental conditions in which the coolant (light water) transcends its saturation temperature, turning into vapor, a homogeneous mixture is used to represent the two-phase flow, and so the coolant channel's energy equation is described using enthalpy. Owing to the fact that temperature and enthalpy are used in the physical model, it became impractical to generate a fully coupled method for solving the pertinent equations. Thus, the conjugated heat transfer problem is solved in a segregated manner through the implementation of an iterative method. Finally, as study cases for this paper we present analyses concerning the behavior of the hottest fuel rod in a Pressurized Water Reactor during a shutdown wherein the residual heat removal system is lost (loss of the reactor's coolant pumps). These studies contemplate cases in which the fuel rod's geometry is ideal or curved. Analyses are also performed for two circumstances of positioning of the fuel inside the rod: concentric and eccentric. (author)

  9. Transient heat transfer analysis up to dryout in 3D fuel rods under unideal conditions through the development of a computer code

    International Nuclear Information System (INIS)

    Martins, Rodolfo I.; Affonso, Renato R.W.; Moreira, Maria de Lourdes; Sampaio, Paulo A. B. de

    2017-01-01

    In this paper we analyze a conjugated transient heat transfer problem consisting of a nuclear reactor's fuel rod and its intrinsic coolant channel. Our analysis is made possible through a computer code being developed at the Instituto de Engenharia Nuclear (IEN/CNEN). This code is meant to study the temperature behavior in fuel rods which exhibit deviation from their ideal conditions, that is, rods in which the cladding is deformed or the fuel is dislocated. It is also designed to avoid the use of the computationally expensive Navier-Stokes equations. For these reasons, its physical model has as basis a three-dimensional fuel rod coupled to a one-dimensional coolant channel, which are discretized using the finite element method. Intending to study accidental conditions in which the coolant (light water) transcends its saturation temperature, turning into vapor, a homogeneous mixture is used to represent the two-phase flow, and so the coolant channel's energy equation is described using enthalpy. Owing to the fact that temperature and enthalpy are used in the physical model, it became impractical to generate a fully coupled method for solving the pertinent equations. Thus, the conjugated heat transfer problem is solved in a segregated manner through the implementation of an iterative method. Finally, as study cases for this paper we present analyses concerning the behavior of the hottest fuel rod in a Pressurized Water Reactor during a shutdown wherein the residual heat removal system is lost (loss of the reactor's coolant pumps). These studies contemplate cases in which the fuel rod's geometry is ideal or curved. Analyses are also performed for two circumstances of positioning of the fuel inside the rod: concentric and eccentric. (author)

  10. Fuel-handling machine tests at the Institute for Nuclear Research - Pitesti. Computer and software research and engineering

    International Nuclear Information System (INIS)

    Doca, Cezar; Predescu, Darie; Maiorescu, Oliviu; Dobrescu, Sorin

    2003-01-01

    This poster introduces the fuel-handling machine SCC-MID. This work is part of a very ambitious project that was accomplished with remarkable investment efforts. Material and human resources was spent to build a test stand for fuel handling machines (CANDU system), closely linked to NPP Cernavoda. A challenging goal was also to develop a computer system (hw/sw) designed and engineered to control the test and calibration process of this fuel-handling machine. The design takes care both of the functionality required to correctly control the F/H machine and of the additional functionality required to assist the testing process. How to test the system itself to validate the implemented solutions, how to safely and consistently maintain the data involved, how to manage such a system, how to gradually integrate the system in the whole stand saving time and work already done and solutions already validated were questions we had to find out right solutions. We choose modular solutions both for hardware and software, based on late technologies which in addition permit to achieve the versatility we needed, namely: VME based hw systems running OS9/68k (Unix like real-time multi-user multitasking OS), ISaGRAF (process control application oriented development and run-time software), Hawk (cross-compiler and IDE software for C/C++ software development intended to run on other Motorola based hw), Suretrack (project management software). The system topology implements open system network concepts that permit communication between different sw/hw platforms (OS9/Motorola and ix86/ms-windows based systems) We spent major resources to model the technological processes and test tools like: - real time simulation of the machine behavior while responding to the human commands or to the state changes of other machine parts as a result of other commands or as mechanical interlinks or technological interlocks and presentation of results revealing time related movements; - database

  11. Model-based diagnosis through Structural Analysis and Causal Computation for automotive Polymer Electrolyte Membrane Fuel Cell systems

    Science.gov (United States)

    Polverino, Pierpaolo; Frisk, Erik; Jung, Daniel; Krysander, Mattias; Pianese, Cesare

    2017-07-01

    The present paper proposes an advanced approach for Polymer Electrolyte Membrane Fuel Cell (PEMFC) systems fault detection and isolation through a model-based diagnostic algorithm. The considered algorithm is developed upon a lumped parameter model simulating a whole PEMFC system oriented towards automotive applications. This model is inspired by other models available in the literature, with further attention to stack thermal dynamics and water management. The developed model is analysed by means of Structural Analysis, to identify the correlations among involved physical variables, defined equations and a set of faults which may occur in the system (related to both auxiliary components malfunctions and stack degradation phenomena). Residual generators are designed by means of Causal Computation analysis and the maximum theoretical fault isolability, achievable with a minimal number of installed sensors, is investigated. The achieved results proved the capability of the algorithm to theoretically detect and isolate almost all faults with the only use of stack voltage and temperature sensors, with significant advantages from an industrial point of view. The effective fault isolability is proved through fault simulations at a specific fault magnitude with an advanced residual evaluation technique, to consider quantitative residual deviations from normal conditions and achieve univocal fault isolation.

  12. Computational investigation of the flow field contribution to improve electricity generation in granular activated carbon-assisted microbial fuel cells

    Science.gov (United States)

    Zhao, Lei; Li, Jian; Battaglia, Francine; He, Zhen

    2016-11-01

    Microbial fuel cells (MFCs) offer an alternative approach to treat wastewater with less energy input and direct electricity generation. To optimize MFC anodic performance, adding granular activated carbon (GAC) has been proved to be an effective way, most likely due to the enlarged electrode surface for biomass attachment and improved mixing of the flow field. The impact of a flow field on the current enhancement within a porous anode medium (e.g., GAC) has not been well understood before, and thus is investigated in this study by using mathematical modeling of the multi-order Butler-Volmer equation with computational fluid dynamics (CFD) techniques. By comparing three different CFD cases (without GAC, with GAC as a nonreactive porous medium, and with GAC as a reactive porous medium), it is demonstrated that adding GAC contributes to a uniform flow field and a total current enhancement of 17%, a factor that cannot be neglected in MFC design. However, in an actual MFC operation, this percentage could be even higher because of the microbial competition and energy loss issues within a porous medium. The results of the present study are expected to help with formulating strategies to optimize MFC with a better flow pattern design.

  13. Notion Of Artificial Labs Slow Global Warming And Advancing Engine Studies Perspectives On A Computational Experiment On Dual-Fuel Compression-Ignition Engine Research

    Directory of Open Access Journals (Sweden)

    Tonye K. Jack

    2017-06-01

    Full Text Available To appreciate clean energy applications of the dual-fuel internal combustion engine D-FICE with pilot Diesel fuel to aid public policy formulation in terms of present and future benefits to the modern transportation stationary power and promotion of oil and gas green- drilling the brief to an engine research team was to investigate the feasible advantages of dual-fuel compression-ignition engines guided by the following concerns i Sustainable fuel and engine power delivery ii The requirements for fuel flexibility iii Low exhausts emissions and environmental pollution iv Achieving low specific fuel consumption and economy for maximum power v The comparative advantages over the conventional Diesel engines vi Thermo-economic modeling and analysis for the optimal blend as basis for a benefitcost evaluation Planned in two stages for reduced cost and fast turnaround of results - initial preliminary stage with basic simple models and advanced stage with more detailed complex modeling. The paper describes a simplified MATLAB based computational experiment predictive model for the thermodynamic combustion and engine performance analysis of dual-fuel compression-ignition engine studies operating on the theoretical limited-pressure cycle with several alternative fuel-blends. Environmental implications for extreme temperature moderation are considered by finite-time thermodynamic modeling for maximum power with predictions for pollutants formation and control by reaction rates kinetics analysis of systematic reduced plausible coupled chemistry models through the NCN reaction pathway for the gas-phase reactions classes of interest. Controllable variables for engine-out pollutants emissions reduction and in particular NOx elimination are identified. Verifications and Validations VampV through Performance Comparisons were made using a clinical approach in selection of StrokeBore ratios greater-than and equal-to one amp88051 low-to-high engine speeds and medium

  14. Improvement of Computer Codes Used for Fuel Behaviour Simulation (FUMEX-III). Report of a Coordinated Research Project 2008-2012

    International Nuclear Information System (INIS)

    2013-03-01

    It is fundamental to the future of nuclear power that reactors can be run safely and economically to compete with other forms of power generation. As a consequence, it is essential to develop the understanding of fuel performance and to embody that knowledge in codes to provide best estimate predictions of fuel behaviour. This in turn leads to a better understanding of fuel performance, a reduction in operating margins, flexibility in fuel management and improved operating economics. The IAEA has therefore embarked on a series of programmes addressing different aspects of fuel behaviour modelling with the following objectives: - To assess the maturity and prediction capabilities of fuel performance codes, and to support interaction and information exchange between countries with code development and application needs (FUMEX series); - To build a database of well defined experiments suitable for code validation in association with the OECD Nuclear Energy Agency (OECD/NEA); - To transfer a mature fuel modelling code to developing countries, to support teams in these countries in their efforts to adapt the code to the requirements of particular reactors, and to provide guidance on applying the code to reactor operation and safety assessments; - To provide guidelines for code quality assurance, code licensing and code application to fuel licensing. This report describes the results of the coordinated research project on the ''Improvement of computer codes used for fuel behaviour simulation (FUMEX-III)''. This programme was initiated in 2008 and completed in 2012. It followed previous programmes on fuel modelling: D-COM 1982-1984, FUMEX 1993-1996 and FUMEX-II 2002-2006. The participants used a mixture of data derived from commercial and experimental irradiation histories, in particular data designed to investigate the mechanical interactions occurring in fuel during normal, transient and severe transient operation. All participants carried out calculations on priority

  15. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    International Nuclear Information System (INIS)

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-01-01

    Highlights: • The burnup of irradiated AGR-1 TRISO fuel was analyzed using gamma spectrometry. • The burnup of irradiated AGR-1 TRISO fuel was also analyzed using mass spectrometry. • Agreement between experimental results and neutron physics simulations was excellent. - Abstract: AGR-1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR-1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non-destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR-1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs-137 activity and the other based on the ratio of Cs-134 and Cs-137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA (fissions per initial heavy metal atom) for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can be determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP-MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma

  16. Modeling optimizes PEM fuel cell durability using three-dimensional multi-phase computational fluid dynamics model

    OpenAIRE

    Maher A.R. Sadiq Al-Baghdadi

    2010-01-01

    Damage mechanisms in a proton exchange membrane (PEM) fuel cell are accelerated by mechanical stresses arising during fuel cell assembly (bolt assembling), and the stresses arise during fuel cell running, because it consists of the materials with different thermal expansion and swelling coefficients. Therefore, in order to acquire a complete understanding of the damage mechanisms in the membrane and gas diffusion layers, mechanical response under steady-state hygro-thermal stresses should be ...

  17. Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Victor [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Kendrick, Brian K. [Theoretical Division (T-1, MS B221), Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Walter, Daniel [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Manera, Annalisa, E-mail: manera@umich.edu [Department of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Boulv, Ann Arbor, MI (United States); Secker, Jeffrey [Westinghouse Electric Company Nuclear Fuel Division, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-04-01

    In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations.

  18. Computational fluid dynamics (CFD) analysis of the combustion process of a leather residuals gasification fuel gas: influence of fuel moisture content

    Energy Technology Data Exchange (ETDEWEB)

    Antonietti, Anderson Jose; Beskow, Arthur Bortolin; Silva, Cristiano Vitorino da [Universidade Regional Integrada do Alto Uruguai e das Missoes (URI), Erechim, RS (Brazil)], E-mails: arthur@uricer.edu.br, mlsperb@unisinos.br; Indrusiak, Maria Luiza Sperb [Universidade do Vale do Rio dos Sinos (UNISINOS), Sao Leopoldo, RS (Brazil)], E-mail: cristiano@uricer.edu.br

    2010-07-01

    This work presents a numerical study of the combustion process of leather residuals gasification gas, aiming the improvement of the process efficiency, considering different concentrations of water on the gas. The heating produced in this combustion process can be used to generation of thermal and/or electrical energy, for use at the leather industrial plant. However, the direct burning of this leather-residual-gas into the chambers is not straightforward. The alternative in development consists in processing this leather residuals by gasification or pyrolysis, separating the volatiles and products of incomplete combustion, for after use as fuel in a boiler. At these processes, different quantities of water can be used, resulting at different levels of moisture content in this fuel gas. This humidity can affect significantly the burning of this fuel, producing unburnt gases, as the carbon monoxide, or toxic gases as NOx, which must have their production minimized on the process, with the purpose of reducing the emission of pollutants to the atmosphere. Other environment-harmful-gases, remaining of the chemical treatment employed at leather manufacture, as cyanide, and hydrocarbons as toluene, must burn too, and the moisture content has influence on it. At this way, to increase understanding of the influence of moisture in the combustion process, it was made a numerical investigation study of reacting flow in the furnace, evaluating the temperature field, the chemical species concentration fields, flow mechanics and heat transfer at the process. The commercial CFD code CFX Ansys Inc. was used. Considering different moisture contents in the fuel used on the combustion process, with this study was possible to achieve the most efficient burning operation parameters, with improvement of combustion efficiency, and reduction of environmental harmful gases emissions. It was verified that the different moisture contents in the fuel gas demand different operation conditions

  19. EXTENDCHAIN: a package of computer programs for calculating the buildup of heavy metals, fission products, and activation products in reactor fuel elements

    International Nuclear Information System (INIS)

    Robertson, M.W.

    1977-01-01

    Design of HTGR recycle and refabrication facilities requires a detailed knowledge of the concentrations of around 400 nuclides which are segregated into four different fuel particle types. The EXTENDCHAIN package of computer programs and the supporting input data files were created to provide an efficient method for calculating the 1600 different concentrations required. The EXTENDCHAIN code performs zero-dimensional nuclide burnup, decay, and activation calculations in nine energy groups for up to 108 nuclides per run. Preparation and handling of the input and output for the sixteen EXTENDCHAIN runs required to produce the desired data are the most time consuming tasks in the computation of the spent fuel element composition. The EXTENDCHAIN package of computer programs contains four codes to aid in the preparation and handling of these data. Most of the input data such as cross sections, decay constants, and the nuclide interconnection scheme will not change when calculating new cases. These data were developed for the life cycle of a typical HTGR and stored on archive tapes for future use. The fuel element composition for this typical HTGR life has been calculated and the results for an equilibrium recycle reload are presented. 12 figures, 7 tables

  20. Reduced-order computational model in nonlinear structural dynamics for structures having numerous local elastic modes in the low-frequency range. Application to fuel assemblies

    International Nuclear Information System (INIS)

    Batou, A.; Soize, C.; Brie, N.

    2013-01-01

    Highlights: • A ROM of a nonlinear dynamical structure is built with a global displacements basis. • The reduced order model of fuel assemblies is accurate and of very small size. • The shocks between grids of a row of seven fuel assemblies are computed. -- Abstract: We are interested in the construction of a reduced-order computational model for nonlinear complex dynamical structures which are characterized by the presence of numerous local elastic modes in the low-frequency band. This high modal density makes the use of the classical modal analysis method not suitable. Therefore the reduced-order computational model is constructed using a basis of a space of global displacements, which is constructed a priori and which allows the nonlinear dynamical response of the structure observed on the stiff part to be predicted with a good accuracy. The methodology is applied to a complex industrial structure which is made up of a row of seven fuel assemblies with possibility of collisions between grids and which is submitted to a seismic loading

  1. Reduced-order computational model in nonlinear structural dynamics for structures having numerous local elastic modes in the low-frequency range. Application to fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Batou, A., E-mail: anas.batou@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Soize, C., E-mail: christian.soize@univ-paris-est.fr [Université Paris-Est, Laboratoire Modélisation et Simulation Multi Echelle, MSME UMR 8208 CNRS, 5 bd Descartes, 77454 Marne-la-Vallee (France); Brie, N., E-mail: nicolas.brie@edf.fr [EDF R and D, Département AMA, 1 avenue du général De Gaulle, 92140 Clamart (France)

    2013-09-15

    Highlights: • A ROM of a nonlinear dynamical structure is built with a global displacements basis. • The reduced order model of fuel assemblies is accurate and of very small size. • The shocks between grids of a row of seven fuel assemblies are computed. -- Abstract: We are interested in the construction of a reduced-order computational model for nonlinear complex dynamical structures which are characterized by the presence of numerous local elastic modes in the low-frequency band. This high modal density makes the use of the classical modal analysis method not suitable. Therefore the reduced-order computational model is constructed using a basis of a space of global displacements, which is constructed a priori and which allows the nonlinear dynamical response of the structure observed on the stiff part to be predicted with a good accuracy. The methodology is applied to a complex industrial structure which is made up of a row of seven fuel assemblies with possibility of collisions between grids and which is submitted to a seismic loading.

  2. COMPUTING

    CERN Multimedia

    M. Kasemann

    Overview In autumn the main focus was to process and handle CRAFT data and to perform the Summer08 MC production. The operational aspects were well covered by regular Computing Shifts, experts on duty and Computing Run Coordination. At the Computing Resource Board (CRB) in October a model to account for service work at Tier 2s was approved. The computing resources for 2009 were reviewed for presentation at the C-RRB. The quarterly resource monitoring is continuing. Facilities/Infrastructure operations Operations during CRAFT data taking ran fine. This proved to be a very valuable experience for T0 workflows and operations. The transfers of custodial data to most T1s went smoothly. A first round of reprocessing started at the Tier-1 centers end of November; it will take about two weeks. The Computing Shifts procedure was tested full scale during this period and proved to be very efficient: 30 Computing Shifts Persons (CSP) and 10 Computing Resources Coordinators (CRC). The shift program for the shut down w...

  3. COMPUTING

    CERN Multimedia

    I. Fisk

    2011-01-01

    Introduction CMS distributed computing system performed well during the 2011 start-up. The events in 2011 have more pile-up and are more complex than last year; this results in longer reconstruction times and harder events to simulate. Significant increases in computing capacity were delivered in April for all computing tiers, and the utilisation and load is close to the planning predictions. All computing centre tiers performed their expected functionalities. Heavy-Ion Programme The CMS Heavy-Ion Programme had a very strong showing at the Quark Matter conference. A large number of analyses were shown. The dedicated heavy-ion reconstruction facility at the Vanderbilt Tier-2 is still involved in some commissioning activities, but is available for processing and analysis. Facilities and Infrastructure Operations Facility and Infrastructure operations have been active with operations and several important deployment tasks. Facilities participated in the testing and deployment of WMAgent and WorkQueue+Request...

  4. COMPUTING

    CERN Multimedia

    P. McBride

    The Computing Project is preparing for a busy year where the primary emphasis of the project moves towards steady operations. Following the very successful completion of Computing Software and Analysis challenge, CSA06, last fall, we have reorganized and established four groups in computing area: Commissioning, User Support, Facility/Infrastructure Operations and Data Operations. These groups work closely together with groups from the Offline Project in planning for data processing and operations. Monte Carlo production has continued since CSA06, with about 30M events produced each month to be used for HLT studies and physics validation. Monte Carlo production will continue throughout the year in the preparation of large samples for physics and detector studies ramping to 50 M events/month for CSA07. Commissioning of the full CMS computing system is a major goal for 2007. Site monitoring is an important commissioning component and work is ongoing to devise CMS specific tests to be included in Service Availa...

  5. COMPUTING

    CERN Multimedia

    M. Kasemann

    Overview During the past three months activities were focused on data operations, testing and re-enforcing shift and operational procedures for data production and transfer, MC production and on user support. Planning of the computing resources in view of the new LHC calendar in ongoing. Two new task forces were created for supporting the integration work: Site Commissioning, which develops tools helping distributed sites to monitor job and data workflows, and Analysis Support, collecting the user experience and feedback during analysis activities and developing tools to increase efficiency. The development plan for DMWM for 2009/2011 was developed at the beginning of the year, based on the requirements from the Physics, Computing and Offline groups (see Offline section). The Computing management meeting at FermiLab on February 19th and 20th was an excellent opportunity discussing the impact and for addressing issues and solutions to the main challenges facing CMS computing. The lack of manpower is particul...

  6. COMPUTING

    CERN Multimedia

    I. Fisk

    2013-01-01

    Computing activity had ramped down after the completion of the reprocessing of the 2012 data and parked data, but is increasing with new simulation samples for analysis and upgrade studies. Much of the Computing effort is currently involved in activities to improve the computing system in preparation for 2015. Operations Office Since the beginning of 2013, the Computing Operations team successfully re-processed the 2012 data in record time, not only by using opportunistic resources like the San Diego Supercomputer Center which was accessible, to re-process the primary datasets HTMHT and MultiJet in Run2012D much earlier than planned. The Heavy-Ion data-taking period was successfully concluded in February collecting almost 500 T. Figure 3: Number of events per month (data) In LS1, our emphasis is to increase efficiency and flexibility of the infrastructure and operation. Computing Operations is working on separating disk and tape at the Tier-1 sites and the full implementation of the xrootd federation ...

  7. RAGRAF: a computer code for calculating temperature distributions in multi-pin fuel assemblies in a stagnant gas atmosphere

    International Nuclear Information System (INIS)

    Eastham, A.

    1979-02-01

    A method of calculating the temperature distribution in a cross-section of a multi-pin nuclear reactor fuel assembly has been computerised. It utilises the thermal radiation interchange between individual fuel pins in either a square or triangular pitched lattice. A stagnant gas atmosphere within the fuel assembly is assumed which inhibits natural convection but permits thermal conduction between adjacent fuel pins. no restriction is placed upon the shape of wrapper used, but its temperature must always be uniform. RAGRAF has great flexibility because of the many options it provides. Although, essentially, it is a transient code, steady state solutions may be readily identified from successive temperature prints. An enclosure for the assembly wrapper is available, to be included or discarded at will during transient calculations. outside the limit of the assembly wrapper, any type or combination of heat transfer mode may be included. Transient variations in boundary temperature may be included if required. (author)

  8. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  9. SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

    International Nuclear Information System (INIS)

    Gulden, W.

    1977-08-01

    The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

  10. Programmer's guide for the CC3 computer models of the concept for disposal of Canada's nuclear fuel waste

    International Nuclear Information System (INIS)

    Dougan, K.D.

    1996-11-01

    Atomic Energy of Canada Limited (AECL) is assessing a concept for disposing of CANDU reactor fuel waste in a vault deep in plutonic rock of the Canadian Shield. A computer program called the Systems Variability Analysis Code (SYVAC) was developed as an analytical tool for the postclosure (long-term) assessment of the concept, and for environmental assessments of other systems. SYVAC3, the third generation of the code, is an executive program that directs repeated simulation of the disposal system, which is described by the CC3 (Canadian Concept, generation 3) model. The CC3 model is comprised of the disposal vault submodel, the local geosphere submodel and the biosphere submodel. The CC3 Proarammer's Guide describes the programming philosophy and programming conventions not covered in the project standards. The guide includes a description of the overall logic for the CC3 vault, geosphere, and biosphere submodels. Each of the CC3 submodels is also isolated from the other two submbdels to create autonomous or 'stand-alone' submodels. The techniques used to isolate a CC3 submodel, and in particular to determine the submodells input and output data interface, are described. Structure charts are provided for the CC3 model and stand-alone submodels. This guide is meant as a companion document to the CC3 User's Manual. This guide does not describe how to use the CC3 software. The user should consult the CC3 User's Manual to determine how to configure, compile, link, and run the CC3 source code, as well as how to modify the data in the input files. It is intended that the CC3 code version CC305 be executed with SYVAC3 version SV309 and the Modelling Algorithm Library (ML3) version ML303, both developed for the assessment of the concept. SYVAC3-CC3-ML3 (also referred to as 'SC3') can be run on any platform containing an ANSI FORTRAN 77 compliant compiler. Recommended hardware environments are specified in the CC3 User's Manual. (author)

  11. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation; Estimativa da queima espacial do combustivel de um reator multiproposito por simulacao computacional

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  12. Development of a Computer Program for an Analysis of the Logistics and Transportation Costs of the PWR Spent Fuels in Korea

    International Nuclear Information System (INIS)

    Cha, Jeong Hun; Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won

    2009-01-01

    It is expected that a substantial amount of spent fuels will be transported from the four nuclear power plant (NPP) sites in Korea to a hypothetical centralized interim storage facility or a final repository in the near future. The cost for the transportation is proportional to the amount of spent fuels. In this paper, a cost estimation program is developed based on the conceptual design of a transportation system and a logistics analysis. Using the developed computer program, named as CASK, the minimum capacity of a centralized interim storage facility (CISF) and the transportation cost for PWR spent fuels are calculated. The PWR spent fuels are transported from 4 NPP sites to a final repository (FR) via the CISF. Since NPP sites and the CISF are located along the coast, a sea-transportation is considered and a road-transportation is considered between the CISF and the FR. The result shows that the minimum capacity of the interim storage facility is 15,000 MTU

  13. Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

    International Nuclear Information System (INIS)

    Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

    2008-07-01

    The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations

  14. Starting Point, Keys and Milestones of a Computer Code for the Simulation of the Behaviour of a Nuclear Fuel Rod

    Directory of Open Access Journals (Sweden)

    Armando C. Marino

    2011-01-01

    Full Text Available The BaCo code (“Barra Combustible” was developed at the Atomic Energy National Commission of Argentina (CNEA for the simulation of nuclear fuel rod behaviour under irradiation conditions. We present in this paper a brief description of the code and the strategy used for the development, improvement, enhancement, and validation of a BaCo during the last 30 years. “Extreme case analysis”, parametric (or sensitivity, probabilistic (or statistic analysis plus the analysis of the fuel performance (full core analysis are the tools developed in the structure of BaCo in order to improve the understanding of the burnup extension in the Atucha I NPP, and the design of advanced fuel elements as CARA and CAREM. The 3D additional tools of BaCo can enhance the understanding of the fuel rod behaviour, the fuel design, and the safety margins. The modular structure of the BaCo code and its detailed coupling of thermo-mechanical and irradiation-induced phenomena make it a powerful tool for the prediction of the influence of material properties on the fuel rod performance and integrity.

  15. COMPUTING

    CERN Multimedia

    I. Fisk

    2010-01-01

    Introduction It has been a very active quarter in Computing with interesting progress in all areas. The activity level at the computing facilities, driven by both organised processing from data operations and user analysis, has been steadily increasing. The large-scale production of simulated events that has been progressing throughout the fall is wrapping-up and reprocessing with pile-up will continue. A large reprocessing of all the proton-proton data has just been released and another will follow shortly. The number of analysis jobs by users each day, that was already hitting the computing model expectations at the time of ICHEP, is now 33% higher. We are expecting a busy holiday break to ensure samples are ready in time for the winter conferences. Heavy Ion An activity that is still in progress is computing for the heavy-ion program. The heavy-ion events are collected without zero suppression, so the event size is much large at roughly 11 MB per event of RAW. The central collisions are more complex and...

  16. COMPUTING

    CERN Multimedia

    M. Kasemann P. McBride Edited by M-C. Sawley with contributions from: P. Kreuzer D. Bonacorsi S. Belforte F. Wuerthwein L. Bauerdick K. Lassila-Perini M-C. Sawley

    Introduction More than seventy CMS collaborators attended the Computing and Offline Workshop in San Diego, California, April 20-24th to discuss the state of readiness of software and computing for collisions. Focus and priority were given to preparations for data taking and providing room for ample dialog between groups involved in Commissioning, Data Operations, Analysis and MC Production. Throughout the workshop, aspects of software, operating procedures and issues addressing all parts of the computing model were discussed. Plans for the CMS participation in STEP’09, the combined scale testing for all four experiments due in June 2009, were refined. The article in CMS Times by Frank Wuerthwein gave a good recap of the highly collaborative atmosphere of the workshop. Many thanks to UCSD and to the organizers for taking care of this workshop, which resulted in a long list of action items and was definitely a success. A considerable amount of effort and care is invested in the estimate of the comput...

  17. COMPUTING

    CERN Multimedia

    P. McBride

    It has been a very active year for the computing project with strong contributions from members of the global community. The project has focused on site preparation and Monte Carlo production. The operations group has begun processing data from P5 as part of the global data commissioning. Improvements in transfer rates and site availability have been seen as computing sites across the globe prepare for large scale production and analysis as part of CSA07. Preparations for the upcoming Computing Software and Analysis Challenge CSA07 are progressing. Ian Fisk and Neil Geddes have been appointed as coordinators for the challenge. CSA07 will include production tests of the Tier-0 production system, reprocessing at the Tier-1 sites and Monte Carlo production at the Tier-2 sites. At the same time there will be a large analysis exercise at the Tier-2 centres. Pre-production simulation of the Monte Carlo events for the challenge is beginning. Scale tests of the Tier-0 will begin in mid-July and the challenge it...

  18. COMPUTING

    CERN Multimedia

    M. Kasemann

    Introduction During the past six months, Computing participated in the STEP09 exercise, had a major involvement in the October exercise and has been working with CMS sites on improving open issues relevant for data taking. At the same time operations for MC production, real data reconstruction and re-reconstructions and data transfers at large scales were performed. STEP09 was successfully conducted in June as a joint exercise with ATLAS and the other experiments. It gave good indication about the readiness of the WLCG infrastructure with the two major LHC experiments stressing the reading, writing and processing of physics data. The October Exercise, in contrast, was conducted as an all-CMS exercise, where Physics, Computing and Offline worked on a common plan to exercise all steps to efficiently access and analyze data. As one of the major results, the CMS Tier-2s demonstrated to be fully capable for performing data analysis. In recent weeks, efforts were devoted to CMS Computing readiness. All th...

  19. COMPUTING

    CERN Multimedia

    I. Fisk

    2011-01-01

    Introduction It has been a very active quarter in Computing with interesting progress in all areas. The activity level at the computing facilities, driven by both organised processing from data operations and user analysis, has been steadily increasing. The large-scale production of simulated events that has been progressing throughout the fall is wrapping-up and reprocessing with pile-up will continue. A large reprocessing of all the proton-proton data has just been released and another will follow shortly. The number of analysis jobs by users each day, that was already hitting the computing model expectations at the time of ICHEP, is now 33% higher. We are expecting a busy holiday break to ensure samples are ready in time for the winter conferences. Heavy Ion The Tier 0 infrastructure was able to repack and promptly reconstruct heavy-ion collision data. Two copies were made of the data at CERN using a large CASTOR disk pool, and the core physics sample was replicated ...

  20. COMPUTING

    CERN Multimedia

    I. Fisk

    2012-01-01

    Introduction Computing continued with a high level of activity over the winter in preparation for conferences and the start of the 2012 run. 2012 brings new challenges with a new energy, more complex events, and the need to make the best use of the available time before the Long Shutdown. We expect to be resource constrained on all tiers of the computing system in 2012 and are working to ensure the high-priority goals of CMS are not impacted. Heavy ions After a successful 2011 heavy-ion run, the programme is moving to analysis. During the run, the CAF resources were well used for prompt analysis. Since then in 2012 on average 200 job slots have been used continuously at Vanderbilt for analysis workflows. Operations Office As of 2012, the Computing Project emphasis has moved from commissioning to operation of the various systems. This is reflected in the new organisation structure where the Facilities and Data Operations tasks have been merged into a common Operations Office, which now covers everything ...

  1. COMPUTING

    CERN Multimedia

    M. Kasemann

    CCRC’08 challenges and CSA08 During the February campaign of the Common Computing readiness challenges (CCRC’08), the CMS computing team had achieved very good results. The link between the detector site and the Tier0 was tested by gradually increasing the number of parallel transfer streams well beyond the target. Tests covered the global robustness at the Tier0, processing a massive number of very large files and with a high writing speed to tapes.  Other tests covered the links between the different Tiers of the distributed infrastructure and the pre-staging and reprocessing capacity of the Tier1’s: response time, data transfer rate and success rate for Tape to Buffer staging of files kept exclusively on Tape were measured. In all cases, coordination with the sites was efficient and no serious problem was found. These successful preparations prepared the ground for the second phase of the CCRC’08 campaign, in May. The Computing Software and Analysis challen...

  2. COMPUTING

    CERN Multimedia

    I. Fisk

    2010-01-01

    Introduction The first data taking period of November produced a first scientific paper, and this is a very satisfactory step for Computing. It also gave the invaluable opportunity to learn and debrief from this first, intense period, and make the necessary adaptations. The alarm procedures between different groups (DAQ, Physics, T0 processing, Alignment/calibration, T1 and T2 communications) have been reinforced. A major effort has also been invested into remodeling and optimizing operator tasks in all activities in Computing, in parallel with the recruitment of new Cat A operators. The teams are being completed and by mid year the new tasks will have been assigned. CRB (Computing Resource Board) The Board met twice since last CMS week. In December it reviewed the experience of the November data-taking period and could measure the positive improvements made for the site readiness. It also reviewed the policy under which Tier-2 are associated with Physics Groups. Such associations are decided twice per ye...

  3. COMPUTING

    CERN Multimedia

    M. Kasemann

    Introduction More than seventy CMS collaborators attended the Computing and Offline Workshop in San Diego, California, April 20-24th to discuss the state of readiness of software and computing for collisions. Focus and priority were given to preparations for data taking and providing room for ample dialog between groups involved in Commissioning, Data Operations, Analysis and MC Production. Throughout the workshop, aspects of software, operating procedures and issues addressing all parts of the computing model were discussed. Plans for the CMS participation in STEP’09, the combined scale testing for all four experiments due in June 2009, were refined. The article in CMS Times by Frank Wuerthwein gave a good recap of the highly collaborative atmosphere of the workshop. Many thanks to UCSD and to the organizers for taking care of this workshop, which resulted in a long list of action items and was definitely a success. A considerable amount of effort and care is invested in the estimate of the co...

  4. A feasibility study on direct methanol fuel cells for laptop computers based on a cost comparison with lithium-ion batteries

    International Nuclear Information System (INIS)

    Wee, Jung-Ho

    2007-01-01

    This paper compares the total cost of direct methanol fuel cell (DMFC) and lithium (Li)-ion battery systems when applied as the power supply for laptop computers in the Korean environment. The average power output and operational time of the laptop computers were assumed to be 20 W and 3000 h, respectively. Considering the status of their technologies and with certain conditions assumed, the total costs were calculated to be US$140 for the Li-ion battery and US$362 for DMFC. The manufacturing costs of the DMFC and Li-ion battery systems were calculated to be $16.65 W -1 and $0.77 W h -1 , and the energy consumption costs to be $0.00051 W h -1 and $0.00032 W h -1 , respectively. The higher fuel consumption cost of the DMFC system was due to the methanol (MeOH) crossover loss. Therefore, the requirements for DMFCs to be able to compete with Li-ion batteries in terms of energy cost include reducing the crossover level to at an order magnitude of -9 and the MeOH price to under $0.5 kg -1 . Under these conditions, if the DMFC manufacturing cost could be reduced to $6.30 W -1 , then the DMFC system would become at least as competitive as the Li-ion battery system for powering laptop computers in Korea. (author)

  5. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays

    International Nuclear Information System (INIS)

    Benedetti, R.L.; Lords, L.V.; Kiser, D.M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage

  6. FRESCO-II: A computer program for analysis of fission product release from spherical HTGR-fuel elements in irradiation and annealing experiments

    International Nuclear Information System (INIS)

    Krohn, H.; Finken, R.

    1983-06-01

    The modular computer code FRESCO has been developed to describe the mechanism of fission product release from a HTGR-Core under accident conditions. By changing some program modules it has been extended to take into account the transport phenomena (i.e. recoil) too, which only occur under reactor operating conditions and during the irradiation experiments. For this report, the release of cesium and strontium from three HTGR-fuel elements has been evaluated and compared with the experimental data. The results show that the measured release can be described by the considered models. (orig.) [de

  7. COMPUTING

    CERN Multimedia

    2010-01-01

    Introduction Just two months after the “LHC First Physics” event of 30th March, the analysis of the O(200) million 7 TeV collision events in CMS accumulated during the first 60 days is well under way. The consistency of the CMS computing model has been confirmed during these first weeks of data taking. This model is based on a hierarchy of use-cases deployed between the different tiers and, in particular, the distribution of RECO data to T1s, who then serve data on request to T2s, along a topology known as “fat tree”. Indeed, during this period this model was further extended by almost full “mesh” commissioning, meaning that RECO data were shipped to T2s whenever possible, enabling additional physics analyses compared with the “fat tree” model. Computing activities at the CMS Analysis Facility (CAF) have been marked by a good time response for a load almost evenly shared between ALCA (Alignment and Calibration tasks - highest p...

  8. COMPUTING

    CERN Multimedia

    Contributions from I. Fisk

    2012-01-01

    Introduction The start of the 2012 run has been busy for Computing. We have reconstructed, archived, and served a larger sample of new data than in 2011, and we are in the process of producing an even larger new sample of simulations at 8 TeV. The running conditions and system performance are largely what was anticipated in the plan, thanks to the hard work and preparation of many people. Heavy ions Heavy Ions has been actively analysing data and preparing for conferences.  Operations Office Figure 6: Transfers from all sites in the last 90 days For ICHEP and the Upgrade efforts, we needed to produce and process record amounts of MC samples while supporting the very successful data-taking. This was a large burden, especially on the team members. Nevertheless the last three months were very successful and the total output was phenomenal, thanks to our dedicated site admins who keep the sites operational and the computing project members who spend countless hours nursing the...

  9. COMPUTING

    CERN Multimedia

    M. Kasemann

    Introduction A large fraction of the effort was focused during the last period into the preparation and monitoring of the February tests of Common VO Computing Readiness Challenge 08. CCRC08 is being run by the WLCG collaboration in two phases, between the centres and all experiments. The February test is dedicated to functionality tests, while the May challenge will consist of running at all centres and with full workflows. For this first period, a number of functionality checks of the computing power, data repositories and archives as well as network links are planned. This will help assess the reliability of the systems under a variety of loads, and identifying possible bottlenecks. Many tests are scheduled together with other VOs, allowing the full scale stress test. The data rates (writing, accessing and transfer¬ring) are being checked under a variety of loads and operating conditions, as well as the reliability and transfer rates of the links between Tier-0 and Tier-1s. In addition, the capa...

  10. COMPUTING

    CERN Multimedia

    Matthias Kasemann

    Overview The main focus during the summer was to handle data coming from the detector and to perform Monte Carlo production. The lessons learned during the CCRC and CSA08 challenges in May were addressed by dedicated PADA campaigns lead by the Integration team. Big improvements were achieved in the stability and reliability of the CMS Tier1 and Tier2 centres by regular and systematic follow-up of faults and errors with the help of the Savannah bug tracking system. In preparation for data taking the roles of a Computing Run Coordinator and regular computing shifts monitoring the services and infrastructure as well as interfacing to the data operations tasks are being defined. The shift plan until the end of 2008 is being put together. User support worked on documentation and organized several training sessions. The ECoM task force delivered the report on “Use Cases for Start-up of pp Data-Taking” with recommendations and a set of tests to be performed for trigger rates much higher than the ...

  11. COMPUTING

    CERN Multimedia

    P. MacBride

    The Computing Software and Analysis Challenge CSA07 has been the main focus of the Computing Project for the past few months. Activities began over the summer with the preparation of the Monte Carlo data sets for the challenge and tests of the new production system at the Tier-0 at CERN. The pre-challenge Monte Carlo production was done in several steps: physics generation, detector simulation, digitization, conversion to RAW format and the samples were run through the High Level Trigger (HLT). The data was then merged into three "Soups": Chowder (ALPGEN), Stew (Filtered Pythia) and Gumbo (Pythia). The challenge officially started when the first Chowder events were reconstructed on the Tier-0 on October 3rd. The data operations teams were very busy during the the challenge period. The MC production teams continued with signal production and processing while the Tier-0 and Tier-1 teams worked on splitting the Soups into Primary Data Sets (PDS), reconstruction and skimming. The storage sys...

  12. COMPUTING

    CERN Multimedia

    I. Fisk

    2013-01-01

    Computing operation has been lower as the Run 1 samples are completing and smaller samples for upgrades and preparations are ramping up. Much of the computing activity is focusing on preparations for Run 2 and improvements in data access and flexibility of using resources. Operations Office Data processing was slow in the second half of 2013 with only the legacy re-reconstruction pass of 2011 data being processed at the sites.   Figure 1: MC production and processing was more in demand with a peak of over 750 Million GEN-SIM events in a single month.   Figure 2: The transfer system worked reliably and efficiently and transferred on average close to 520 TB per week with peaks at close to 1.2 PB.   Figure 3: The volume of data moved between CMS sites in the last six months   The tape utilisation was a focus for the operation teams with frequent deletion campaigns from deprecated 7 TeV MC GEN-SIM samples to INVALID datasets, which could be cleaned up...

  13. COMPUTING

    CERN Multimedia

    I. Fisk

    2012-01-01

      Introduction Computing activity has been running at a sustained, high rate as we collect data at high luminosity, process simulation, and begin to process the parked data. The system is functional, though a number of improvements are planned during LS1. Many of the changes will impact users, we hope only in positive ways. We are trying to improve the distributed analysis tools as well as the ability to access more data samples more transparently.  Operations Office Figure 2: Number of events per month, for 2012 Since the June CMS Week, Computing Operations teams successfully completed data re-reconstruction passes and finished the CMSSW_53X MC campaign with over three billion events available in AOD format. Recorded data was successfully processed in parallel, exceeding 1.2 billion raw physics events per month for the first time in October 2012 due to the increase in data-parking rate. In parallel, large efforts were dedicated to WMAgent development and integrati...

  14. Computational sensitivity study of spray dispersion and mixing on the fuel properties in a gas turbine combustor

    Energy Technology Data Exchange (ETDEWEB)

    Grosshans, Holger; Szász, Robert-Zoltán [Division of Fluid Mechanics, Lund University (Sweden); Cao, Le [Key Laboratory for Aerosol-Cloud-Precipitation of China Meteorological Administration, Nanjing University of Information Science and Technology, Nanjing (China); Fuchs, Laszlo, E-mail: holger.grosshans@uclouvain.be [Department of Mechanics, KTH, Stockholm (Sweden)

    2017-04-15

    A swirl stabilized gas turbine burner has been simulated in order to assess the effects of the fuel properties on spray dispersion and fuel–air mixing. The properties under consideration include fuel surface tension, viscosity and density. The turbulence of the gas phase is modeled applying the methodology of large eddy simulation whereas the dispersed liquid phase is described by Lagrangian particle tracking. The exchange of mass, momentum and energy between the two phases is accounted for by two-way coupling. Bag and stripping breakup regimes are considered for secondary droplet breakup, using the Reitz–Diwakar and the Taylor analogy breakup models. Moreover, a model for droplet evaporation is included. The results reveal a high sensitivity of the spray structure to variations of all investigated parameters. In particular, a decrease in the surface tension or the fuel viscosity, or an increase in the fuel density, lead to less stable liquid structures. As a consequence, smaller droplets are generated and the overall spray surface area increases, leading to faster evaporation and mixing. Furthermore, with the trajectories of the small droplets being strongly influenced by aerodynamic forces (and less by their own inertia), the spray is more affected by the turbulent structures of the gaseous phase and the spray dispersion is enhanced. (paper)

  15. COBRA - 3C/KFKI: a digital computer program for steady and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

    International Nuclear Information System (INIS)

    Vigassy, J.; Kovacs, L.M.

    1977-11-01

    COBRA-3C/KFKI is a digital computer program for the CDC-3300 computer in FORTRAN language. The program is a revised version of the original COBRA-3C code. The code calculates steady-state and transient flow and enthalpy transport in rod-bundle nuclear fuel elements in both boiling and nonboiling conditions. The mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The program neglects sonic velocity propagation but allows for a temporal and spatial acceleration of the diversion crossflow in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate and exit pressure. (D.P.)

  16. COMPUTING

    CERN Multimedia

    I. Fisk

    2011-01-01

    Introduction The Computing Team successfully completed the storage, initial processing, and distribution for analysis of proton-proton data in 2011. There are still a variety of activities ongoing to support winter conference activities and preparations for 2012. Heavy ions The heavy-ion run for 2011 started in early November and has already demonstrated good machine performance and success of some of the more advanced workflows planned for 2011. Data collection will continue until early December. Facilities and Infrastructure Operations Operational and deployment support for WMAgent and WorkQueue+Request Manager components, routinely used in production by Data Operations, are provided. The GlideInWMS and components installation are now deployed at CERN, which is added to the GlideInWMS factory placed in the US. There has been new operational collaboration between the CERN team and the UCSD GlideIn factory operators, covering each others time zones by monitoring/debugging pilot jobs sent from the facto...

  17. Analysis of Trends and Emerging Technologies in Water Electrolysis Research Based on a Computational Method: A Comparison with Fuel Cell Research

    Directory of Open Access Journals (Sweden)

    Takaya Ogawa

    2018-02-01

    Full Text Available Water electrolysis for hydrogen production has received increasing attention, especially for accumulating renewable energy. Here, we comprehensively reviewed all water electrolysis research areas through computational analysis, using a citation network to objectively detect emerging technologies and provide interdisciplinary data for forecasting trends. The results show that all research areas increase their publication counts per year, and the following two areas are particularly increasing in terms of number of publications: “microbial electrolysis” and “catalysts in an alkaline water electrolyzer (AWE and in a polymer electrolyte membrane water electrolyzer (PEME.”. Other research areas, such as AWE and PEME systems, solid oxide electrolysis, and the whole renewable energy system, have recently received several review papers, although papers that focus on specific technologies and are cited frequently have not been published within the citation network. This indicates that these areas receive attention, but there are no novel technologies that are the center of the citation network. Emerging technologies detected within these research areas are presented in this review. Furthermore, a comparison with fuel cell research is conducted because water electrolysis is the reverse reaction to fuel cells, and similar technologies are employed in both areas. Technologies that are not transferred between fuel cells and water electrolysis are introduced, and future water electrolysis trends are discussed.

  18. Dynamic Thermal Loads and Cooling Requirements Calculations for V ACs System in Nuclear Fuel Processing Facilities Using Computer Aided Energy Conservation Models

    International Nuclear Information System (INIS)

    EL Fawal, M.M.; Gadalla, A.A.; Taher, B.M.

    2010-01-01

    In terms of nuclear safety, the most important function of ventilation air conditioning (VAC) systems is to maintain safe ambient conditions for components and structures important to safety inside the nuclear facility and to maintain appropriate working conditions for the plant's operating and maintenance staff. As a part of a study aimed to evaluate the performance of VAC system of the nuclear fuel cycle facility (NFCF) a computer model was developed and verified to evaluate the thermal loads and cooling requirements for different zones of fuel processing facility. The program is based on transfer function method (TFM) and it is used to calculate the dynamic heat gain by various multilayer walls constructions and windows hour by hour at any orientation of the building. The developed model was verified by comparing the obtained calculated results of the solar heat gain by a given building with the corresponding calculated values using finite difference method (FDM) and total equivalent temperature different method (TETD). As an example the developed program is used to calculate the cooling loads of the different zones of a typical nuclear fuel facility the results showed that the cooling capacities of the different cooling units of each zone of the facility meet the design requirements according to safety regulations in nuclear facilities.

  19. CALIPSO - a computer code for the calculation of fluiddynamics, thermohydraulics and changes of geometry in failing fuel elements of a fast breeder reactor

    International Nuclear Information System (INIS)

    Kedziur, F.

    1982-07-01

    The computer code CALIPSO was developed for the calculation of a hypothetical accident in an LMFBR (Liquid Metal Fast Breeder Reactor), where the failure of fuel pins is assumed. It calculates two-dimensionally the thermodynamics, fluiddynamics and changes in geometry of a single fuel pin and its coolant channel in a time period between failure of the pin and a state, at which the geometry is nearly destroyed. The determination of temperature profiles in the fuel pin cladding and the channel wall make it possible to take melting and freezing processes into account. Further features of CALIPSO are the variable channel cross section in order to model disturbances of the channel geometry as well as the calculation of two velocity fields including the consideration of virtual mass effects. The documented version of CALIPSO is especially suited for the calculation of the SIMBATH experiments carried out at the Kernforschungszentrum Karlsruhe, which simulate the above-mentioned accident. The report contains the complete documentation of the CALIPSO code: the modeling of the geometry, the equations used, the structure of the code and the solution procedure as well as the instructions for use with an application example. (orig.) [de

  20. Fuel pattern recognition device

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1995-01-01

    The device of the present invention monitors normal fuel exchange upon fuel exchanging operation carried out in a reactor of a nuclear power plant. Namely, a fuel exchanger is movably disposed to the upper portion of the reactor and exchanges fuels. An exclusive computer receives operation signals of the fuel exchanger during operation as inputs, and outputs reactor core fuel pattern information signals to a fuel arrangement diagnosis device. An underwater television camera outputs image signals of a fuel pattern in the reactor core to an image processing device. If there is any change in the image signals for the fuel pattern as a result of the fuel exchange operation of the fuel exchanger, the image processing device outputs the change as image signals to the fuel pattern diagnosis device. The fuel pattern diagnosis device compares the pattern information signals from the exclusive computer with the image signals from the image processing device, to diagnose the result of the fuel exchange operation performed by the fuel exchanger and inform the diagnosis by means of an image display. (I.S.)

  1. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  2. Neutron-photon energy deposition in CANDU reactor fuel channels: a comparison of modelling techniques using ANISN and MCNP computer codes

    International Nuclear Information System (INIS)

    Bilanovic, Z.; McCracken, D.R.

    1994-12-01

    In order to assess irradiation-induced corrosion effects, coolant radiolysis and the degradation of the physical properties of reactor materials and components, it is necessary to determine the neutron, photon, and electron energy deposition profiles in the fuel channels of the reactor core. At present, several different computer codes must be used to do this. The most recent, advanced and versatile of these is the latest version of MCNP, which may be capable of replacing all the others. Different codes have different assumptions and different restrictions on the way they can model the core physics and geometry. This report presents the results of ANISN and MCNP models of neutron and photon energy deposition. The results validate the use of MCNP for simplified geometrical modelling of energy deposition by neutrons and photons in the complex geometry of the CANDU reactor fuel channel. Discrete ordinates codes such as ANISN were the benchmark codes used in previous work. The results of calculations using various models are presented, and they show very good agreement for fast-neutron energy deposition. In the case of photon energy deposition, however, some modifications to the modelling procedures had to be incorporated. Problems with the use of reflective boundaries were solved by either including the eight surrounding fuel channels in the model, or using a boundary source at the bounding surface of the problem. Once these modifications were incorporated, consistent results between the computer codes were achieved. Historically, simple annular representations of the core were used, because of the difficulty of doing detailed modelling with older codes. It is demonstrated that modelling by MCNP, using more accurate and more detailed geometry, gives significantly different and improved results. (author). 9 refs., 12 tabs., 20 figs

  3. COMPUTING

    CERN Multimedia

    M. Kasemann

    CMS relies on a well functioning, distributed computing infrastructure. The Site Availability Monitoring (SAM) and the Job Robot submission have been very instrumental for site commissioning in order to increase availability of more sites such that they are available to participate in CSA07 and are ready to be used for analysis. The commissioning process has been further developed, including "lessons learned" documentation via the CMS twiki. Recently the visualization, presentation and summarizing of SAM tests for sites has been redesigned, it is now developed by the central ARDA project of WLCG. Work to test the new gLite Workload Management System was performed; a 4 times increase in throughput with respect to LCG Resource Broker is observed. CMS has designed and launched a new-generation traffic load generator called "LoadTest" to commission and to keep exercised all data transfer routes in the CMS PhE-DEx topology. Since mid-February, a transfer volume of about 12 P...

  4. Analysis on specific nuclear data for reactors physics computations applied to CANDU reactors using thorium-based fuels

    International Nuclear Information System (INIS)

    Visan, Iuliana E.

    2010-01-01

    The purpose of this work is to analyze the evaluated nuclear data from ENDF libraries IAEA69 (69 energy groups library) and IAEA172 (172 energy groups library), respectively, in WIMS library format and to represent neutron fission yield, absorption and fission cross-section dependence for 233 Uranium, 232 Thorium isotopes and some actinides of interest on the incident energy. Our interest for these two isotopes is mainly based on the importance of 233 Uranium as 'fissile nucleus' in Thorium-Uranium fuel cycle. Nowadays, nuclear data evaluation for the actinides generated in Thorium-Uranium fuel cycle is seen as a world-wide priority. The fissile nucleus, 233 Uranium 'plays' the same function in Thorium-Uranium fuel cycle as the 235 Uranium in 'the classic' Uranium-Plutonium fuel cycle. As opposed to natural Uranium which contains 0.7 % of the fissile isotope 235 Uranium, natural Thorium doesn't contain fissile isotopes, being composed entirely by the fertile isotope 232 Thorium. Graphical evolutions of interest parameters versus the incident energy are presented. Our interest was also to observe the behavior of these nuclear data for fast, resonance and thermal energy groups, respectively. The ENDF nuclear data libraries are constantly up-dated, so that we can observe an improvement of the IAEA172 library, which disposes of evaluated nuclear data at higher energies (about 20 MeV), as opposed to IAEA69 library (which includes evaluated nuclear data below 10 MeV). Based on our graphical representation, a good agreement between the considered libraries has been observed, sustaining nuclear data validity. (authors)

  5. Systems-level computational modeling demonstrates fuel selection switching in high capacity running and low capacity running rats

    Science.gov (United States)

    Qi, Nathan R.

    2018-01-01

    High capacity and low capacity running rats, HCR and LCR respectively, have been bred to represent two extremes of running endurance and have recently demonstrated disparities in fuel usage during transient aerobic exercise. HCR rats can maintain fatty acid (FA) utilization throughout the course of transient aerobic exercise whereas LCR rats rely predominantly on glucose utilization. We hypothesized that the difference between HCR and LCR fuel utilization could be explained by a difference in mitochondrial density. To test this hypothesis and to investigate mechanisms of fuel selection, we used a constraint-based kinetic analysis of whole-body metabolism to analyze transient exercise data from these rats. Our model analysis used a thermodynamically constrained kinetic framework that accounts for glycolysis, the TCA cycle, and mitochondrial FA transport and oxidation. The model can effectively match the observed relative rates of oxidation of glucose versus FA, as a function of ATP demand. In searching for the minimal differences required to explain metabolic function in HCR versus LCR rats, it was determined that the whole-body metabolic phenotype of LCR, compared to the HCR, could be explained by a ~50% reduction in total mitochondrial activity with an additional 5-fold reduction in mitochondrial FA transport activity. Finally, we postulate that over sustained periods of exercise that LCR can partly overcome the initial deficit in FA catabolic activity by upregulating FA transport and/or oxidation processes. PMID:29474500

  6. PRE-CASKETSS: an input data generation computer program for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-12-01

    A computer program PRE-CASKETSS has been developed for the purpose of input data generation for thermal and structural analysis computer code system CASKETSS (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of PRE-CASKETSS are as follow; (1) Function of input data generation for thermal and structural analysis computer programs is provided in the program. (2) Two- and three-dimensional mesh generation for finite element and finite difference programs are available in the program. (3) The capacity of the material input data generation are provided in the program. (4) The boundary conditions, the load conditions and the initial conditions are capable in the program. (5) This computer program operate both the time shearing system and the batch system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  7. Quantitative analysis of the fission product distribution in a damaged fuel assembly using gamma-spectrometry and computed tomography for the Phébus FPT3 test

    International Nuclear Information System (INIS)

    Biard, B.

    2013-01-01

    Highlights: • The FP quantitative distribution in the fuel bundle is measured by gamma-spectrometry. • The FP location is obtained with emission tomograms and other experiment results. • X-ray tomograms provide the material and density mapping of the degraded bundle. • The self-attenuation may then be computed for each isotope at its key line energy. • Results are consistent with other FPT3 measurements, with acceptable uncertainties. -- Abstract: The international Phébus FP programme, initiated in 1988 by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), in cooperation with the European Commission (EC) and with financial support from USNRC, Canada, PSI/HSK (Switzerland), Japan and Korea, was aimed at studying severe accident phenomena: the fuel degradation, the release of fission products (FPs) and their transport through the reactor coolant system to the containment building. The FPT3 test, conducted in 2004, was the last of the five light water reactor core meltdown accident tests performed on irradiated fuel rods. After the experiment, the test device was recovered and analysed through a full set of non-destructive examinations performed over the fuel bundle zone, including gamma-scanning, gamma emission tomography, X-ray radiography and X-ray transmission tomography. The gamma-scanning was specifically devoted to the location, identification and amount quantification of the FPs remaining in the bundle. Since the fuel bundle became highly degraded during the experiment, the geometry was different at each level examined, and did not correspond to the well-known initial state. The self-attenuation of the test device and consequently the efficiency correction could then not be estimated by classical means that need to know the geometry of the object. Using the results of the other non-destructive examinations, specific computational tools and methods have therefore been developed to compute the self-attenuation of the bundle

  8. Quantitative analysis of the fission product distribution in a damaged fuel assembly using gamma-spectrometry and computed tomography for the Phébus FPT3 test

    Energy Technology Data Exchange (ETDEWEB)

    Biard, B., E-mail: bruno.biard@irsn.fr

    2013-09-15

    Highlights: • The FP quantitative distribution in the fuel bundle is measured by gamma-spectrometry. • The FP location is obtained with emission tomograms and other experiment results. • X-ray tomograms provide the material and density mapping of the degraded bundle. • The self-attenuation may then be computed for each isotope at its key line energy. • Results are consistent with other FPT3 measurements, with acceptable uncertainties. -- Abstract: The international Phébus FP programme, initiated in 1988 by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), in cooperation with the European Commission (EC) and with financial support from USNRC, Canada, PSI/HSK (Switzerland), Japan and Korea, was aimed at studying severe accident phenomena: the fuel degradation, the release of fission products (FPs) and their transport through the reactor coolant system to the containment building. The FPT3 test, conducted in 2004, was the last of the five light water reactor core meltdown accident tests performed on irradiated fuel rods. After the experiment, the test device was recovered and analysed through a full set of non-destructive examinations performed over the fuel bundle zone, including gamma-scanning, gamma emission tomography, X-ray radiography and X-ray transmission tomography. The gamma-scanning was specifically devoted to the location, identification and amount quantification of the FPs remaining in the bundle. Since the fuel bundle became highly degraded during the experiment, the geometry was different at each level examined, and did not correspond to the well-known initial state. The self-attenuation of the test device and consequently the efficiency correction could then not be estimated by classical means that need to know the geometry of the object. Using the results of the other non-destructive examinations, specific computational tools and methods have therefore been developed to compute the self-attenuation of the bundle

  9. Vehicle Design Evaluation Program (VDEP). A computer program for weight sizing, economic, performance and mission analysis of fuel-conservative aircraft, multibodied aircraft and large cargo aircraft using both JP and alternative fuels

    Science.gov (United States)

    Oman, B. H.

    1977-01-01

    The NASA Langley Research Center vehicle design evaluation program (VDEP-2) was expanded by (1) incorporating into the program a capability to conduct preliminary design studies on subsonic commercial transport type aircraft using both JP and such alternate fuels as hydrogen and methane;(2) incorporating an aircraft detailed mission and performance analysis capability; and (3) developing and incorporating an external loads analysis capability. The resulting computer program (VDEP-3) provides a preliminary design tool that enables the user to perform integrated sizing, structural analysis, and cost studies on subsonic commercial transport aircraft. Both versions of the VDEP-3 Program which are designated preliminary Analysis VDEP-3 and detailed Analysis VDEP utilize the same vehicle sizing subprogram which includes a detailed mission analysis capability, as well as a geometry and weight analysis for multibodied configurations.

  10. Computations between metallocalix(4)arene host and a series of four oil-based fuel pollutant guests

    International Nuclear Information System (INIS)

    Pathak, D.A.; Street, N.C.

    2006-01-01

    Calculations using PM3 and mechanics methods on metallocalix(4)arene hosts (1-10) and substituted dibenzothiophene guests (A-D), which are generally known as oil-based fuel pollutants, show that host-guest formation is energetically favored. Calculations have been carried out for both 1/1 and 1/4 ratios of host/guest. There is no direct bonding between the metal center of the host and the sulfur of the guest in the host-guest complex. Sterically hundered dibenzothiophene guests show similar energies to the unhindered analogs. For calix(4)arenas (5-10) in partial cone conformations and having hydrogen rather than p-tert-butyl groups on the wide rim, host-guest formation occurs within the narrow rim rather than the wide rim. Host-guest association appears to occur via Pie-Pie interactions between host and guest phenyl groups rather than via metal-sulfur bonding. The study has importance especially in oil refining to obtain environmentally safe fuel oils and help supramolecular chemists in designing and synthesizing more sophisticated host molecules for the removal of sulfur from crude oil / refinery oil. (author)

  11. Resolving the three-dimensional microstructure of polymer electrolyte fuel cell electrodes using nanometer-scale X-ray computed tomography

    Energy Technology Data Exchange (ETDEWEB)

    Epting, William K.; Gelb, Jeff; Litster, Shawn

    2012-02-08

    The electrodes of a polymer electrolyte fuel cell (PEFC) are composite porous layers consisting of carbon and platinum nanoparticles and a polymer electrolyte binder. The proper composition and arrangement of these materials for fast reactant transport and high electrochemical activity is crucial to achieving high performance, long lifetimes, and low costs. Here, the microstructure of a PEFC electrode using nanometer-scale X-ray computed tomography (nano-CT) with a resolution of 50 nm is investigated. The nano-CT instrument obtains this resolution for the low-atomic-number catalyst support and binder using a combination of a Fresnel zone plate objective and Zernike phase contrast imaging. High-resolution, non-destructive imaging of the three-dimensional (3D) microstructures provides important new information on the size and form of the catalyst particle agglomerates and pore spaces. Transmission electron microscopy (TEM) and mercury intrusion porosimetry (MIP) is applied to evaluate the limits of the resolution and to verify the 3D reconstructions. The computational reconstructions and size distributions obtained with nano-CT can be used for evaluating electrode preparation, performing pore-scale simulations, and extracting effective morphological parameters for large-scale computational models. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  12. POST-CASKETSS: a graphic computer program for thermal and structural analysis of nuclear fuel shipping casks

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-12-01

    A computer program POST-CASKETSS has been developed for the purpose of calculation result representation for thermal and structural analysis computer code system CASKETSS (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of POST-CASKETSS are as follows; (1) Function of calculation result representation for thermal and structural analysis computer programs is provided in the program. (2) Two and three dimensional graphic representation for finite element and finite difference programs are available in the program. (3) The capacity of graphics of geometry, temperature contor and temperature-time curve are provided for thermal analysis. (4) The capacity of graphics of geometry, deformation, stress contor, displacement-time curve, velocity-time curve, acceleration-time curve, stress-time curve, force-time curve and moment-time curve are provided for structural analysis. (5) This computer program operates both the time shearing system and the batch system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

  13. The simulation of the alternate turbopump development high pressure oxygen and fuel turbopumps for the space shuttle main engine using the Shaberth computer program

    Science.gov (United States)

    Mcdonald, Gary H.

    1988-01-01

    The Space Shuttle Main Engine (SSME) is basically comprised of a combustion chamber and nozzle, high and low pressure oxygen turbopumps and high and low pressure fuel turbopumps. In the current configuration, the high pressure fuel (HPTFP) and high pressure oxygen turbopumps (HPOTP) have experienced a history of ball bearing wear. The wear problem can be attributed to numerous factors including the hydrodynamic axial and radial loads caused by the flow of liquid oxygen and liquid hydrogen through the turbopump impellers and turbine. Also, friction effects between the rolling elements, races, and cage can create thermally induced bearing geometry changes. To alleviate some of the current configuration problems, an alternate turbopump development (ATD) was proposed. However, the ATD HPOTP and HPTFP are constrained to operate interchangeably with the current turbopumps, thus, the operation conditions must be similar. The ATD configuration features a major change in bearings used to support the integrated shaft, impeller, and turbine system. A single ball and single roller will replace the pump-end and turbine and duplex ball bearings. The Shaft-Bearing-Thermal (SHABERTH) computer code was used to model the ATD HPOTP and ATD HPFTP configurations. A two bearing model was used to simulate the HPOTP and HPFTP bearings and shaft geometry. From SHABERTH, a comparison of bearing reaction loads, frictional heat generation rates, and Hertz contact stresses will be attempted with analysis at the 109 percent and 65 percent power levels.

  14. User`s manual for the CC3 computer models of the concept for disposal of Canada`s nuclear fuel waste

    Energy Technology Data Exchange (ETDEWEB)

    Dougan, K D; Wojciechowski, L C

    1995-06-01

    Atomic Energy of Canada Limited (AECL) is assessing a concept for disposing of CANDU reactor fuel waste in a vault deep in plutonic rock of the Canadian Shield. A computer program called the Systems Variability Analysis Code (SYVAC) has been developed as an analytical tool for the postclosure (long-term) assessment of the concept, and for environmental assessments of other systems. SYVAC3, the third generation of the code, is an executive program that directs repeated simulation of the disposal system, which is represented by the CC3 (Canadian Concept, generation 3) models comprising a design-specific vault, a site-specific geosphere and a biosphere typical of the Canadian Shield. (author). 23 refs., 7 tabs., 21 figs.

  15. BASIC program to compute uranium density and void volume fraction in laboratory-scale uranium silicide aluminum dispersion plate-type fuel

    International Nuclear Information System (INIS)

    Ugajin, Mitsuhiro

    1991-05-01

    BASIC program simple and easy to operate has been developed to compute uranium density and void volume fraction for laboratory-scale uranium silicide aluminum dispersion plate-type fuel, so called miniplate. An example of the result of calculation is given in order to demonstrate how the calculated void fraction correlates with the microstructural distribution of the void in a miniplate prepared in our laboratory. The program is also able to constitute data base on important parameters for miniplates from experimentally-determined values of density, weight of each constituent and dimensions of miniplates. Utility programs pertinent to the development of the BASIC program are also given which run in the popular MS-DOS environment. All the source lists are attached and brief description for each program is made. (author)

  16. TAO2000 V2 computer-assisted force feedback tele-manipulators used as maintenance and production tools at the AREVA NC-La Hague fuel recycling plant

    International Nuclear Information System (INIS)

    Geffard, Franck; Garrec, Philippe; Piolain, Gerard; Brudieu, Marie-Anne; Thro, Jean-Francois; Coudray, Alain; Lelann, Eric

    2012-01-01

    During a 15-year joint research program, French Atomic Energy Agency Interactive Robotics Laboratory (CEA LIST) and AREVA have developed several remote operation devices, also called tele-robots. Some of them are now commonly used for maintenance operations at the AREVA NC (Nuclear Cycle) La Hague reprocessing plant. Since the first maintenance operation in 2005, several other successful interventions have been realized using the industrial MA23/RX170 tele-manipulation system. Moreover, since 2010, the through-the-wall tele-robot named MT200 TAO based on the slave arm of the MSM MT200 (La Calhene TM ), has been evaluated in an active production cell at the AREVA NC La Hague fuel recycling plant. Although these evaluations are ongoing, the positive results obtained have led to an update and industrialization program. All these developments are based on the same generic control platform, called TAO2000 V2. TAO2000 V2 is the second release of the CEA LIST core software platform dedicated to computer aided force-feedback tele-operation (TAO is the French acronym for computer aided tele-operation). This paper presents all these developments resulting from the joint research program CEA LIST/AREVA. The TAO2000 V2 controller is first detailed, and then two maintenance operations using the industrial robot RX170 are presented: the removal of the nuclear fuel dissolver wheel rollers and the cleanup of the dissolver wheel inter-bucket spaces. Finally, the new MT200 TAO system and its evaluations at the AREVA NC La Hague facilities are discussed. (authors)

  17. Computational and Experimental Studies of Microstructure-Scale Porosity in Metallic Fuels for Improved Gas Swelling Behavior

    Energy Technology Data Exchange (ETDEWEB)

    Mllett, Paul [Univ. of Arkansas, Fayetteville, AR (United States); McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Deo, Chaitanya [Georgia Inst. of Technology, Atlanta, GA (United States); Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2018-01-29

    This proposal will investigate the stability of bimodal pore size distributions in metallic uranium and uranium-zirconium alloys during sintering and re-sintering annealing treatments. The project will utilize both computational and experimental approaches. The computational approach includes both Molecular Dynamics simulations to determine the self-diffusion coefficients in pure U and U-Zr alloys in single crystals, grain boundaries, and free surfaces, as well as calculations of grain boundary and free surface interfacial energies. Phase-field simulations using MOOSE will be conducted to study pore and grain structure evolution in microstructures with bimodal pore size distributions. Experiments will also be performed to validate the simulations, and measure the time-dependent densification of bimodal porous compacts.

  18. QUEIMAP: a computer routine for punctual analysis of a nuclear fuel depletion with accumulation of fission fragments

    International Nuclear Information System (INIS)

    Couto, R.T.

    1990-01-01

    QUEIMAP is a computer routine for burnup calculation, composed of five FORTRAN-77 subroutines. Its objective is to solve depletion equation of four radionuclides conversion chain, U238, U235, Th232, as well as fission fragments equations. In this paper the burnup is considered punctual and evolutioned under cross section. It presents the solution algorithms employed by QUEIMAP, the validation of its results and the way of use it. (M.I.)

  19. Inter-comparison of Computer Codes for TRISO-based Fuel Micro-Modeling and Performance Assessment

    International Nuclear Information System (INIS)

    Boer, Brian; Keun Jo, Chang; Wu, Wen; Ougouag, Abderrafi M.; McEachren, Donald; Venneri, Francesco

    2010-01-01

    The Next Generation Nuclear Plant (NGNP), the Deep Burn Pebble Bed Reactor (DB-PBR) and the Deep Burn Prismatic Block Reactor (DB-PMR) are all based on fuels that use TRISO particles as their fundamental constituent. The TRISO particle properties include very high durability in radiation environments, hence the designs reliance on the TRISO to form the principal barrier to radioactive materials release. This durability forms the basis for the selection of this fuel type for applications such as Deep Bun (DB), which require exposures up to four times those expected for light water reactors. It follows that the study and prediction of the durability of TRISO particles must be carried as part of the safety and overall performance characterization of all the designs mentioned above. Such evaluations have been carried out independently by the performers of the DB project using independently developed codes. These codes, PASTA, PISA and COPA, incorporate models for stress analysis on the various layers of the TRISO particle (and of the intervening matrix material for some of them), model for fission products release and migration then accumulation within the SiC layer of the TRISO particle, just next to the layer, models for free oxygen and CO formation and migration to the same location, models for temperature field modeling within the various layers of the TRISO particle and models for the prediction of failure rates. All these models may be either internal to the code or external. This large number of models and the possibility of different constitutive data and model formulations and the possibility of a variety of solution techniques makes it highly unlikely that the model would give identical results in the modeling of identical situations. The purpose of this paper is to present the results of an inter-comparison between the codes and to identify areas of agreement and areas that need reconciliation. The inter-comparison has been carried out by the cooperating

  20. Development of a computational model applied to a unitary 144 cm{sup 2} proton exchange membrane fuel cell; Desenvolvimento de um modelo numerico computacional aplicado a uma celula a combustivel unitaria de 144 CM{sup 2} tipo PEM

    Energy Technology Data Exchange (ETDEWEB)

    Robalinho, Eric

    2009-07-01

    This work presents the development of a numerical computer model and methodology to study and design polymeric exchange membrane fuel cell - PEM. For the validation of experimental results, a sequence of routines, appropriate to fit the data obtained in the laboratory, was described. At the computational implementation it was created a new strategy of coupling two 3-dimensional models to satisfy the requirements of the comprehensive model of the fuel cell, including its various geometries and materials, as well as the various physical and chemical processes simulated. To effective assessment of the real cell analogy with numerical model, numerical studies were carried out. Comparisons with values obtained in the literature, characterization of variables through laboratory experiments and estimates from models already tested in the literature were also performed. Regarding the experimental part, a prototype of a fuel cell unit of 144 cm of geometric area was designed, produced and operated at laboratory with the purpose of validating the numerical computer model proposed, with positive results. The results of simulations for the 2D and 3D geometries proposed are presented in the form of polarization curves, highlighting the catalytic layer model based on the geometry of agglomerates. Parametric and sensitivity studies are presented to illustrate the change in performance of the fuel cell studied. The final model is robust and useful as a tool for design and optimization of PEM type fuel cells in a wide range of operating conditions. (author)

  1. Development of a computational model applied to a unitary 144 CM{sup 2} proton exchange membrane fuel cell; Desenvolvimento de um modelo numerico computacional aplicado a uma celula a combustivel unitaria de 144 CM{sup 2} tipo PEM

    Energy Technology Data Exchange (ETDEWEB)

    Robalinho, Eric

    2009-07-01

    This work presents the development of a numerical computer model and methodology to study and design polymeric exchange membrane fuel cell - PEM. For the validation of experimental results, a sequence of routines, appropriate to fit the data obtained in the laboratory, was described. At the computational implementation it was created a new strategy of coupling two 3-dimensional models to satisfy the requirements of the comprehensive model of the fuel cell, including its various geometries and materials, as well as the various physical and chemical processes simulated. To effective assessment of the real cell analogy with numerical model, numerical studies were carried out. Comparisons with values obtained in the literature, characterization of variables through laboratory experiments and estimates from models already tested in the literature were also performed. Regarding the experimental part, a prototype of a fuel cell unit of 144 cm{sup 2} of geometric area was designed, produced and operated at laboratory with the purpose of validating the numerical computer model proposed, with positive results. The results of simulations for the 2D and 3D geometries proposed are presented in the form of polarization curves, highlighting the catalytic layer model based on the geometry of agglomerates. Parametric and sensitivity studies are presented to illustrate the change in performance of the fuel cell studied. The final model is robust and useful as a tool for design and optimization of PEM type fuel cells in a wide range of operating conditions. (author)

  2. 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.

  3. Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Dionne, B.

    2011-01-01

    To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D

  4. Application of X-ray micro-computed tomography on high-speed cavitating diesel fuel flows

    Energy Technology Data Exchange (ETDEWEB)

    Mitroglou, N.; Lorenzi, M.; Gavaises, M. [City University London, School of Mathematics Computer Science and Engineering, London (United Kingdom); Santini, M. [University of Bergamo, Department of Engineering, Bergamo (Italy)

    2016-11-15

    The flow inside a purpose built enlarged single-orifice nozzle replica is quantified using time-averaged X-ray micro-computed tomography (micro-CT) and high-speed shadowgraphy. Results have been obtained at Reynolds and cavitation numbers similar to those of real-size injectors. Good agreement for the cavitation extent inside the orifice is found between the micro-CT and the corresponding temporal mean 2D cavitation image, as captured by the high-speed camera. However, the internal 3D structure of the developing cavitation cloud reveals a hollow vapour cloud ring formed at the hole entrance and extending only at the lower part of the hole due to the asymmetric flow entry. Moreover, the cavitation volume fraction exhibits a significant gradient along the orifice volume. The cavitation number and the needle valve lift seem to be the most influential operating parameters, while the Reynolds number seems to have only small effect for the range of values tested. Overall, the study demonstrates that use of micro-CT can be a reliable tool for cavitation in nozzle orifices operating under nominal steady-state conditions. (orig.)

  5. Application of X-ray micro-computed tomography on high-speed cavitating diesel fuel flows

    Science.gov (United States)

    Mitroglou, N.; Lorenzi, M.; Santini, M.; Gavaises, M.

    2016-11-01

    The flow inside a purpose built enlarged single-orifice nozzle replica is quantified using time-averaged X-ray micro-computed tomography (micro-CT) and high-speed shadowgraphy. Results have been obtained at Reynolds and cavitation numbers similar to those of real-size injectors. Good agreement for the cavitation extent inside the orifice is found between the micro-CT and the corresponding temporal mean 2D cavitation image, as captured by the high-speed camera. However, the internal 3D structure of the developing cavitation cloud reveals a hollow vapour cloud ring formed at the hole entrance and extending only at the lower part of the hole due to the asymmetric flow entry. Moreover, the cavitation volume fraction exhibits a significant gradient along the orifice volume. The cavitation number and the needle valve lift seem to be the most influential operating parameters, while the Reynolds number seems to have only small effect for the range of values tested. Overall, the study demonstrates that use of micro-CT can be a reliable tool for cavitation in nozzle orifices operating under nominal steady-state conditions.

  6. Application of X-ray micro-computed tomography on high-speed cavitating diesel fuel flows

    International Nuclear Information System (INIS)

    Mitroglou, N.; Lorenzi, M.; Gavaises, M.; Santini, M.

    2016-01-01

    The flow inside a purpose built enlarged single-orifice nozzle replica is quantified using time-averaged X-ray micro-computed tomography (micro-CT) and high-speed shadowgraphy. Results have been obtained at Reynolds and cavitation numbers similar to those of real-size injectors. Good agreement for the cavitation extent inside the orifice is found between the micro-CT and the corresponding temporal mean 2D cavitation image, as captured by the high-speed camera. However, the internal 3D structure of the developing cavitation cloud reveals a hollow vapour cloud ring formed at the hole entrance and extending only at the lower part of the hole due to the asymmetric flow entry. Moreover, the cavitation volume fraction exhibits a significant gradient along the orifice volume. The cavitation number and the needle valve lift seem to be the most influential operating parameters, while the Reynolds number seems to have only small effect for the range of values tested. Overall, the study demonstrates that use of micro-CT can be a reliable tool for cavitation in nozzle orifices operating under nominal steady-state conditions. (orig.)

  7. 3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

    Directory of Open Access Journals (Sweden)

    Gongyi Yu

    2017-01-01

    Full Text Available A nonuniform distribution of tristructural isotropic (TRISO particles within a high-temperature gas-cooled reactor (HTGR pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT: first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control.

  8. Final report on LDRD project : elucidating performance of proton-exchange-membrane fuel cells via computational modeling with experimental discovery and validation.

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Chao Yang (Pennsylvania State University, University Park, PA); Pasaogullari, Ugur (Pennsylvania State University, University Park, PA); Noble, David R.; Siegel, Nathan P.; Hickner, Michael A.; Chen, Ken Shuang

    2006-11-01

    In this report, we document the accomplishments in our Laboratory Directed Research and Development project in which we employed a technical approach of combining experiments with computational modeling and analyses to elucidate the performance of hydrogen-fed proton exchange membrane fuel cells (PEMFCs). In the first part of this report, we document our focused efforts on understanding water transport in and removal from a hydrogen-fed PEMFC. Using a transparent cell, we directly visualized the evolution and growth of liquid-water droplets at the gas diffusion layer (GDL)/gas flow channel (GFC) interface. We further carried out a detailed experimental study to observe, via direct visualization, the formation, growth, and instability of water droplets at the GDL/GFC interface using a specially-designed apparatus, which simulates the cathode operation of a PEMFC. We developed a simplified model, based on our experimental observation and data, for predicting the onset of water-droplet instability at the GDL/GFC interface. Using a state-of-the-art neutron imaging instrument available at NIST (National Institute of Standard and Technology), we probed liquid-water distribution inside an operating PEMFC under a variety of operating conditions and investigated effects of evaporation due to local heating by waste heat on water removal. Moreover, we developed computational models for analyzing the effects of micro-porous layer on net water transport across the membrane and GDL anisotropy on the temperature and water distributions in the cathode of a PEMFC. We further developed a two-phase model based on the multiphase mixture formulation for predicting the liquid saturation, pressure drop, and flow maldistribution across the PEMFC cathode channels. In the second part of this report, we document our efforts on modeling the electrochemical performance of PEMFCs. We developed a constitutive model for predicting proton conductivity in polymer electrolyte membranes and compared

  9. Computer modeling of obesity links theoretical energetic measures with experimental measures of fuel use for lean and obese men.

    Science.gov (United States)

    Rossow, Heidi A; Calvert, C Chris

    2014-10-01

    The goal of this research was to use a computational model of human metabolism to predict energy metabolism for lean and obese men. The model is composed of 6 state variables representing amino acids, muscle protein, visceral protein, glucose, triglycerides, and fatty acids (FAs). Differential equations represent carbohydrate, amino acid, and FA uptake and output by tissues based on ATP creation and use for both lean and obese men. Model parameterization is based on data from previous studies. Results from sensitivity analyses indicate that model predictions of resting energy expenditure (REE) and respiratory quotient (RQ) are dependent on FA and glucose oxidation rates with the highest sensitivity coefficients (0.6, 0.8 and 0.43, 0.15, respectively, for lean and obese models). Metabolizable energy (ME) is influenced by ingested energy intake with a sensitivity coefficient of 0.98, and a phosphate-to-oxygen ratio by FA oxidation rate and amino acid oxidation rate (0.32, 0.24 and 0.55, 0.65 for lean and obese models, respectively). Simulations of previously published studies showed that the model is able to predict ME ranging from 6.6 to 9.3 with 0% differences between published and model values, and RQ ranging from 0.79 to 0.86 with 1% differences between published and model values. REEs >7 MJ/d are predicted with 6% differences between published and model values. Glucose oxidation increases by ∼0.59 mol/d, RQ increases by 0.03, REE increases by 2 MJ/d, and heat production increases by 1.8 MJ/d in the obese model compared with lean model simulations. Increased FA oxidation results in higher changes in RQ and lower relative changes in REE. These results suggest that because fat mass is directly related to REE and rate of FA oxidation, body fat content could be used as a predictor of RQ. © 2014 American Society for Nutrition.

  10. Development and verification of the LIFE-GCFR computer code for predicting gas-cooled fast-reactor fuel-rod performance

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Billone, M.C.; Rest, J.

    1982-03-01

    The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects

  11. Radiological Risk Assessments for Occupational Exposure at Fuel Fabrication Facility in AlTuwaitha Site Baghdad – Iraq by using RESRAD Computer Code

    Science.gov (United States)

    Ibrahim, Ziadoon H.; Ibrahim, S. A.; Mohammed, M. K.; Shaban, A. H.

    2018-05-01

    The purpose of this study is to evaluate the radiological risks for workers for one year of their activities at Fuel Fabrication Facility (FFF) so as to make the necessary protection to prevent or minimize risks resulted from these activities this site now is under the Iraqi decommissioning program (40). Soil samples surface and subsurface were collected from different positions of this facility and analyzed by gamma rays spectroscopy technique High Purity Germanium detector (HPGe) was used. It was found out admixture of radioactive isotopes (232Th 40K 238U 235U137Cs) according to the laboratory results the highest values were (975758) for 238U (21203) for 235U (218) for 232Th (4046) for 40K and (129) for 137Cs in (Bqkg1) unit. The annual total radiation dose and risks were estimated by using RESRAD (onsite) 70 computer code. The highest total radiation dose was (5617μSv/year) in area that represented by soil sample (S7) and the radiological risks morbidity and mortality (118E02 8661E03) respectively in the same area

  12. Development of a computer program of fast calculation for the pre design of advanced nuclear fuel 10 x 10 for BWR type reactors

    International Nuclear Information System (INIS)

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

    2005-01-01

    In the National Institute of Nuclear Research (ININ) a methodology is developed to optimize the design of cells 10x10 of assemble fuels for reactors of water in boil or BWR. It was proposed a lineal calculation formula based on a coefficients matrix (of the change reason of the relative power due to changes in the enrichment of U-235) for estimate the relative powers by pin of a cell. With this it was developed the computer program of fast calculation named PreDiCeldas. The one which by means of a simple search algorithm allows to minimize the relative power peak maximum of cell or LPPF. This is achieved varying the distribution of U-235 inside the cell, maintaining in turn fixed its average enrichment. The accuracy in the estimation of the relative powers for pin is of the order from 1.9% when comparing it with results of the 'best estimate' HELIOS code. With the PreDiCeldas it was possible, at one minimum time of calculation, to re-design a reference cell diminishing the LPPF, to the beginning of the life, of 1.44 to a value of 1.31. With the cell design with low LPPF is sought to even design cycles but extensive that those reached at the moment in the BWR of the Laguna Verde Central. (Author)

  13. 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.)

  14. Validation of computer codes and modelling methods for giving proof of nuclear saefty of transport and storage of spent VVER-type nuclear fuels. Part 1. Purposes and goals of the project. Final report

    International Nuclear Information System (INIS)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-01-01

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k eff is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [de

  15. 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)

  16. The spent fuel safety experiment

    International Nuclear Information System (INIS)

    Harmms, G.A.; Davis, F.J.; Ford, J.T.

    1995-01-01

    The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

  17. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods; Simulacao com programas computacionais de desempenho do combustivel em regimes permanente e transiente de varetas combustiveis de aco inoxidavel austenitico

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza

    2014-07-01

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  18. 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.

  19. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-01

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor

  20. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-15

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

  1. 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.)

  2. 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.)

  3. Improvement of computer programs 'BAMBOO' and 'ASFRE-IV' for coupling analysis of deformation and thermal-hydraulics in a high burn-up fuel subassembly of fast reactor

    International Nuclear Information System (INIS)

    Uwaba, Tomoyuki; Ohshima, Hiroyuki; Imai, Yasutomo

    2003-04-01

    A simulation system of a deformed fuel subassembly is being developed for the structure integrity of high burn-up wire-spacer-type fuel subassemblies of sodium-cooled fast breeder reactors. This report describes a computer program improvement work for coupling analyses of deformation and thermal-hydraulics in a fuel subassembly as part of the simulation system development. In this work, a function of data conversion as an interface between a bundle deformation analysis program BAMBOO and a thermal hydraulic analysis program ASFRE-IV was incorporated to each program. BAMBOO was improved to accept the coolant temperature data from ASFRE-IV and to offer bundle deformation data to ASFRE-IV. ASFRE-IV was also improved to offer the coolant temperature data to BAMBOO and to obtain the bundle deformation data from BAMBOO. Improved BAMBOO and ASFRE-IV were applied to an analysis of 169-pin bundle for the program verification. It was confirmed that the coupling analysis gave the physically reasonable results on both deformation and thermal hydraulic behaviors in the fuel subassembly. (author)

  4. Vibration of fuel bundles

    International Nuclear Information System (INIS)

    Chen, S.S.

    1975-06-01

    Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code, AMASS, is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension, and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods

  5. 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

  6. 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.)

  7. 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.)

  8. 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)

  9. A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

    International Nuclear Information System (INIS)

    Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A.; Nowak, M.; Pencer, J.; Novog, D.; Buijs, A.

    2015-01-01

    Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg"-"1 [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

  10. SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

    1978-02-01

    The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocity and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.

  11. A feasibility study on the use of the MOOSE computational framework to simulate three-dimensional deformation of CANDU reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, Kyle A., E-mail: Kyle.Gamble@inl.gov [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Williams, Anthony F., E-mail: Tony.Williams@cnl.ca [Canadian Nuclear Laboratories, Fuel and Fuel Channel Safety, 1 Plant Road, Chalk River, Ontario, Canada K0J 1J0 (Canada); Chan, Paul K., E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Chemistry and Chemical Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada); Wowk, Diane, E-mail: Diane.Wowk@rmc.ca [Royal Military College of Canada, Mechanical and Aerospace Engineering, 13 General Crerar Crescent, Kingston, Ontario, Canada K7K 7B4 (Canada)

    2015-11-15

    Highlights: • This is the first demonstration of using the MOOSE framework for modeling CANDU fuel. • Glued and frictionless contact algorithms behave as expected for 2D and 3D cases. • MOOSE accepts and correctly interprets functions of arbitrary form. • 3D deformation calculations accurately compare against analytical solutions. • MOOSE is a viable simulation tool for modeling accident reactor conditions. - Abstract: Horizontally oriented fuel bundles, such as those in CANada Deuterium Uranium (CANDU) reactors present unique modeling challenges. After long irradiation times or during severe transients the fuel elements can laterally deform out of plane due to processes known as bow and sag. Bowing is a thermally driven process that causes the fuel elements to laterally deform when a temperature gradient develops across the diameter of the element. Sagging is a coupled mechanical and thermal process caused by deformation of the fuel pin due to creep mechanisms of the sheathing after long irradiation times and or high temperatures. These out-of-plane deformations can lead to reduced coolant flow and a reduction in coolability of the fuel bundle. In extreme cases element-to-element or element-to-pressure tube contact could occur leading to reduced coolant flow in the subchannels or pressure tube rupture leading to a loss of coolant accident. This paper evaluates the capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at the Idaho National Laboratory to model these deformation mechanisms. The material model capabilities of MOOSE and its ability to simulate contact are also investigated.

  12. Modeling of PHWR fuel elements using FUDA code

    International Nuclear Information System (INIS)

    Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.

    2008-01-01

    The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated

  13. User's guide for FREG-3: a computer program to analyze pellet-cladding gap conductance in accordance with fuel-rod irradiation history

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio; Fujita, Misao; Ishibashi, Akihiro; Otsubo, Naoaki.

    1976-10-01

    The present report describes user's manual for program FREG-3, and provides a general description of the program and instructions of input/output. FREG-3 estimates the temperature distribution in a fuel rod and the stored energy based on the distribution. The temperature distribution is calculated in accordance with fuel-rod irradiation history. Mechanical properties and models in handling specific problems, such as densification and relocation, are optional in the program. The options are to be given by key word. If appropriate options are selected, the program is used not only as a safety evaluation code, but also as a best evaluation code. (auth.)

  14. Development and implementation of computational geometric model for simulation of plate type fuel fabrication process with microspheres dispersed in metallic matrix

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Reis, Sergio C.; Braga, Daniel M.; Santos, Armindo; Ferraz, Wilmar B.

    2005-01-01

    In this report it is presented the development of a geometric model to simulate the plate type fuel fabrication process with fuels microspheres dispersed in metallic matrix, as well as its software implementation. The developed geometric model encloses the steps of pellets pressing and sintering, as well as the plate rolling passes. The model permits the simulation of structures, where the values of the various variables of the fabrication processes can be studied and modified. The following variables were analyzed: microspheres diameters, density of the powder/microspheres mixing, microspheres density, fuel volume fraction, sintering densification, and rolling passes number. In the model implementation, which was codified in DELPHI programming language, systems of structured analysis techniques were utilized. The structures simulated were visualized utilizing the AutoCAD applicative, what permitted to obtain planes sections in diverse directions. The objective of this model is to enable the analysis of the simulated structures and supply information that can help in the improvement of the dispersion microspheres fuel plates fabrication process, now in development at CDTN (Centro de Desenvolvimento da Tecnologia Nuclear) in cooperation with the CTMSP (Centro Tecnologico da Marinha em Sao Paulo). (author)

  15. The state of the art in non destructive testing of nuclear fuel cladding tubes using ultrasonic rotary systems; on line computer and statistics

    International Nuclear Information System (INIS)

    Rauscher, Rudolf

    Nondestructive evaluation of nuclear fuel cladding by ultrasonic tests is described. Ultrasonic transducers for detection of flaws and dimensions are built in a rotary system with a speed of 8000 rpm. The testing system is adapted to a configuration consisting of two microcomputers connected to each other

  16. Novel techniques of oxygen bleeding for polymer electrolyte fuel cells under impure anode feeding and poisoning condition: A computational study using OpenFOAM®

    International Nuclear Information System (INIS)

    Hafttananian, M.; Ramiar, A.; Ranjbar, A.A.

    2016-01-01

    Highlights: • Oxygen bleeding as a treatment of poisoned fuel cells is simulated. • Oxygen bleeding, depended on impurity concentration, has an optimize range. • Three novel techniques of oxygen bleeding are introduced. • The OBME parameter is introduced to compare oxygen bleeding methods. • Pulsating and sinusoidal methods lead to 2.74 and 1.74 effectiveness, respectively. - Abstract: PEM fuel cells fed by steam methane reformate (SMR) and water gas shift reformate (WGSR) are always subjected to catalyst poisoning due to trace amounts of carbon monoxide in the fuel stream. Air or oxygen bleeding are the most common, cost-effective and easy-to-implement methods of mitigation for such conditions. In the present work, important restrictions of oxygen bleeding and disadvantages of continuous constant O_2 bleeding are introduced. In order to simulating the problem, a mathematical model containing reaction kinetics, governing equations, boundary conditions and constitutive relations are presented and according to these, a comprehensive OpenFOAM transient multicomponent solver having the capability of simulating anode side of PEM fuel cell is developed, introduced and validated with existing experimental data. In addition, three novel methods of oxygen bleeding: namely pulsating, exponential and sinusoidal methods are introduced which are practical for both open-end and dead-end fuel cells. A quantitative criterion, Oxygen Bleeding Method Effectiveness (OBME), is also introduced to attain comparability of novel methods with continuous constant O_2 bleeding as common method. Results showed up to more than 5% increase in recovery ratio accompanied with up to 63% save in oxygen consumption.

  17. Application of modified version of SPPS-1 - HEXAB-2DB computer code package for operational analyses of fuel behaviour in WWER-440 reactors at Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kharalampieva, Ts; Stoyanova, I; Antonov, A; Simeonov, T [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Petkov, P [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1994-12-31

    The modified version of SPPS-1 code called SPPS-1-HEXAB-2DB was applied for the purposes of the operational analysis and power peaking factors and reactor core critical parameters predictions of WWER-440s. The results of the calculations performed by the use of SPPS-1-HEXAB-2DB code and the corresponding parameters obtained from experiments at Kozloduy NPP WWER-440s as well as the results of fuel rod power distribution are presented. The method of operation simulation of reactor core with 349 assemblies (Unit 4) and with 313 fuel assemblies and 36 dummy fuel assemblies (Unit 1) is outlined. The modified code calculates not only fuel burnup and Pm-149 and Sm-149 concentrations distributions but also the space distribution of I-135 and Xe-135 concentrations. In this way it makes possible to perform the reactor operation simulation during the immediate periods after the reactor start-up or shut-down and to predict the critical reactor core parameters during transients. The results obtained show that SPPS-1-HEXAB-2DB code describes adequately the reactor core status. The new SPPS-1 code algorithm for estimation of assembly-wise power peaking factors distribution in reactor core is also described. The new code provides an option for checking the correctness of reactor core symmetry. The experience from the use of the modified SPPS-1-HEXAB-2DB code system confirms the provision of improved availability of operational analysis, prediction of Kozloduy NPP WWER-440s safe operations and fuel behaviour estimation. 14 tabs., 4 figs., 5 refs.

  18. 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.)

  19. 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

  20. The fuel performance code future

    International Nuclear Information System (INIS)

    Ronchi, C.; Van de Laar, J.

    1988-01-01

    The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

  1. 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

  2. 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.)

  3. 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

  4. Neutronic calculations with transport and diffusion computer codes for light water moderated critical with UO2 enriched at 4,75% as fuel

    International Nuclear Information System (INIS)

    Sabundjian, G.; Nakata, H.

    1983-02-01

    The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt

  5. 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

  6. A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A., E-mail: sharpejr@mcmaster.ca [McMaster University, Hamilton, ON (Canada); Nowak, M. [McMaster University, Hamilton, ON (Canada); Institut National Polytechnique de Grenoble, Phelma, Grenoble (France); Pencer, J. [McMaster University, Hamilton, ON (Canada); Canadian Nuclear Laboratories, Chalk River, ON, (Canada); Novog, D.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

    2015-07-01

    Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg{sup -1} [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

  7. 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

  8. 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…

  9. 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

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

    International Nuclear Information System (INIS)

    Levine, S.H.

    1980-07-01

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

  11. 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

  12. Fuel Burn Estimation Using Real Track Data

    Science.gov (United States)

    Chatterji, Gano B.

    2011-01-01

    A procedure for estimating fuel burned based on actual flight track data, and drag and fuel-flow models is described. The procedure consists of estimating aircraft and wind states, lift, drag and thrust. Fuel-flow for jet aircraft is determined in terms of thrust, true airspeed and altitude as prescribed by the Base of Aircraft Data fuel-flow model. This paper provides a theoretical foundation for computing fuel-flow with most of the information derived from actual flight data. The procedure does not require an explicit model of thrust and calibrated airspeed/Mach profile which are typically needed for trajectory synthesis. To validate the fuel computation method, flight test data provided by the Federal Aviation Administration were processed. Results from this method show that fuel consumed can be estimated within 1% of the actual fuel consumed in the flight test. Next, fuel consumption was estimated with simplified lift and thrust models. Results show negligible difference with respect to the full model without simplifications. An iterative takeoff weight estimation procedure is described for estimating fuel consumption, when takeoff weight is unavailable, and for establishing fuel consumption uncertainty bounds. Finally, the suitability of using radar-based position information for fuel estimation is examined. It is shown that fuel usage could be estimated within 5.4% of the actual value using positions reported in the Airline Situation Display to Industry data with simplified models and iterative takeoff weight computation.

  13. 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.)

  14. 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

  15. 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.)

  16. Use of an expert system for the data acquisition input of a computer software for materials flux of an irradiated fuel reprocessing plant

    International Nuclear Information System (INIS)

    Ouvrier, N.; Castelli, P.

    1993-01-01

    An irradiated fuel reprocessing plant products purified plutonium and all sorts of wastes. PROBILUS program has been developed to calculate the transfer flux of materials. The data acquisition input being long, an expert system MARCMOD has been realized. PROBILUS was conceived like a standards collection. We wanted to automate the tedious tasks, to decrease the output result delays, and to improve the result quality. That is the reason why an expert system MARCMOD with a graphic interface has been carried out. With this tool the user may describe graphically his process, check the coherence of acquired informations, generate automatically the whole input data of PROBILUS, and change by interaction some predefined data

  17. Computer code system for the R and D of nuclear fuel cycle with fast reactor. 2. Development and application of analytical evaluation system for thermal striping phenomena

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu

    2001-01-01

    Fluid-structure thermal interaction phenomena characterized by stationary random temperature fluctuations, namely thermal striping are observed in the downstream region such as a T-junction piping system of liquid metal fast reactors (LMFRs). Therefore, the piping wall located in the downstream region must be protected against the stationary random thermal process, which might induce high-cycle fatigue. This paper describes the evaluation system based on numerical simulation methods consisting of three thermohydraulics computer programs AQUA, DINUS-3 and THEMIS and of three thermomechanical computer programs BEMSET, FINAS and CANIS, for the thermal striping developed at Japan Nuclear Cycle Development Institute (JNC). Verification results for each computer code and the system are also introduced based on out-of-pile experimental data using water and sodium as working fluids. (author)

  18. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  19. 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

  20. Fuel channel refilling

    International Nuclear Information System (INIS)

    Shoukri, M.; Abdul-Razzak, A.

    1990-04-01

    Analysis of existing data on fuel channel refilling is presented. The analysis focuses on the data obtained using the Stern Laboratories Cold Water Injection Test (CWIT) Facility. The two-fluid model thermal-hydraulics computer code CATHENA is also used to simulate experimental results on fuel channel refilling from both the CWIT and RD-14 facilities. Conclusions related to single and double break tests, including the effect of non-condensible gases, are presented. A set of recommendations is given for further analysis and separate effect experiments. (67 figs., 5 tabs., 24 refs.)

  1. Correction device for fuel positioning value

    International Nuclear Information System (INIS)

    Nakahara, Toshiro.

    1993-01-01

    In a computer for a fuel exchanger for controlling handling of fuels such as loading and unloading based on the data for settling values of fuels installed in a reactor core or fuel pool, data for burnup degree during reactor operation are inputted from a computer computing the reactor output into a correction device for fuel settling values in which fuel irradiation growth rate is calculated to determine a correction value. This makes it unnecessary for sampling measurement of fuel settling values in the reactor core practiced at the same time with the reactor opening and exchange operation for the stored date of settling values in the computer for the fuel exchanger conducted on every time. New data for settling values can be exchanged automatically based on the data for the burnup degree at the same time with the reactor shutdown, which can be conducted easily over the whole number. Accordingly, it is possible to improve reliability and safety of the fuel exchanging operation including the setting for the interlock of the fuel exchanging operation relevant to the fuel settling values, as well as moderate operator's burden. (N.H.)

  2. 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.)

  3. 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.)

  4. 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.)

  5. 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.)

  6. 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.)

  7. 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....

  8. 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

  9. 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

  10. 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.)

  11. 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

  12. Computational and Experimental Study of the Structure of Diffusion Flames of Jet Fuel and Its Surrogates at Pressures up to 40 ATM

    Science.gov (United States)

    2012-11-21

    examination of some of the aromatics show that the model captures well benzene from toluene decomposition in BF, but underpredicts styrene and ethylbenzene ...critical toluene pyrolysis products and stable soot precursors were compared with computational models using two semi-detailed chemical mechanisms... ethylbenzene , which at least one of the mechanisms reproduces quite well. The largest measured species in the incipiently sooting flame is indene, whose

  13. Automated breeder fuel fabrication

    International Nuclear Information System (INIS)

    Goldmann, L.H.; Frederickson, J.R.

    1983-01-01

    The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures

  14. Optical fuel spray measurements

    Energy Technology Data Exchange (ETDEWEB)

    Hillamo, H.

    2011-07-01

    Diesel fuel sprays, including fuel/air mixing and the physics of two-phase jet formation, are discussed in the thesis. The fuel/air mixing strongly affects emissions formation in spray combustion processes where the local combustion conditions dictate the emission formation. This study comprises optical measurements both in pressurized spray test rigs and in a running engine.The studied fuel injection was arranged with a common rail injection system and the injectors were operated with a solenoid-based injection valve. Both marine and heavy-duty diesel engine injectors were used in the study. Optical fuel spray measurements were carried out with a laser-based double-framing camera system. This kind of equipments is usually used for flow field measurements with Particle Image Velocimetry technique (PIV) as well as for backlight imaging. Fundamental fuel spray properties and spray formation were studied in spray test rigs. These measurements involved studies of mixing, atomization, and the flow field. Test rig measurements were used to study the effect of individual injection parameters and component designs. Measurements of the fuel spray flow field, spray penetration, spray tip velocity, spray angle, spray structure, droplet accumulation, and droplet size estimates are shown. Measurement campaign in a running optically accessible large-bore medium-speed engine was also carried out. The results from engine tests were compared with equivalent test rig measurements, as well as computational results, to evaluate the level of understanding of sprays. It was shown that transient spray has an acceleration and a deceleration phase. Successive flow field measurements (PIV) in optically dense diesel spray resulted in local and average velocity data of diesel sprays. Processing fuel spray generates a flow field to surrounding gas and entrainment of surrounding gas into fuel jet was also seen at the sides of the spray. Laser sheet imaging revealed the inner structure of diesel

  15. Preliminary analysis of aircraft fuel systems for use with broadened specification jet fuels

    Science.gov (United States)

    Pasion, A. J.; Thomas, I.

    1977-01-01

    An analytical study was conducted on the use of broadened specification hydrocarbon fuels in present day aircraft. A short range Boeing 727 mission and three long range Boeing 747 missions were used as basis of calculation for one-day-per-year extreme values of fuel loading, airport ambient and altitude ambient temperatures with various seasonal and climatic conditions. Four hypothetical fuels were selected; two high-vapor-pressure fuels with 35 kPa and 70 kPa RVP and two high-freezing-point fuels with -29 C and -18 C freezing points. In-flight fuel temperatures were predicted by Boeing's aircraft fuel tank thermal analyzer computer program. Boil-off rates were calculated for the high vapor pressure fuels and heating/insulation requirements for the high freezing point fuels were established. Possible minor and major heating system modifications were investigated with respect to heat output, performance and economic penalties for the high freezing point fuels.

  16. 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)

  17. 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.

  18. 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

  19. 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.)

  20. Core status computing system

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki.

    1982-01-01

    Purpose: To calculate power distribution, flow rate and the like in the reactor core with high accuracy in a BWR type reactor. Constitution: Total flow rate signals, traverse incore probe (TIP) signals as the neutron detector signals, thermal power signals and pressure signals are inputted into a process computer, where the power distribution and the flow rate distribution in the reactor core are calculated. A function generator connected to the process computer calculates the absolute flow rate passing through optional fuel assemblies using, as variables, flow rate signals from the introduction part for fuel assembly flow rate signals, data signals from the introduction part for the geometrical configuration data at the flow rate measuring site of fuel assemblies, total flow rate signals for the reactor core and the signals from the process computer. Numerical values thus obtained are given to the process computer as correction signals to perform correction for the experimental data. (Moriyama, K.)

  1. 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

  2. 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.)

  3. CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

  4. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    Ayer, J.E.; Clark, A.T.; Loysen, P.; Ballinger, M.Y.; Mishima, J.; Owczarski, P.C.; Gregory, W.S.; Nichols, B.D.

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  5. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 1 presents these data for unirradiated fuel, uranium ore and uranium mill tailings. In Part 2 they have been computed for fuel irradiated to levels of burnup ranging from 140 GJ/kg U to 1150 GJ/kg U. (author)

  6. 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....

  7. 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.)

  8. 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

  9. 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)

  10. 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.)

  11. 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

  12. 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.)

  13. The sphere-PAC fuel code 'SPHERE-3'

    International Nuclear Information System (INIS)

    Wallin, H.

    2000-01-01

    Sphere-PAC fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-PAC fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

  14. The sphere-pac fuel code 'SPHERE-3'

    International Nuclear Information System (INIS)

    Wallin, H.; Nordstroem, L.A.; Hellwig, C.

    2001-01-01

    Sphere-pac fuel is an advanced nuclear fuel, in which the cladding tube is filled with small fuel spheres instead of the more usual fuel pellets. At PSI, the irradiation behaviour of sphere-pac fuel is calculated using the computer code SPHERE-3. The paper describes the present status of the SPHERE-3 code, and some results of the qualification process against experimental data. (author)

  15. Software development assurance for special control computer sets on the base of the Electronica-60M microcomputer in the stations of fuel element nondestructive testing

    International Nuclear Information System (INIS)

    Vasil'kova, A.D.; Grachev, S.N.; Salakatova, L.S.

    1987-01-01

    A special computational control complex (SCCC) including the ''Elektronika-60M'' microcomputer, a device for communication with an object (DCO) based on the typical microprocessing set (TMS) based on LIUS-2 KTS means and an adapter for communication of the ''Elektronika-60M'' microcomputer bus with IK1 TMS bus developed for their conjugation is used for development of nondestructive control post software. An instrumental SCCC including SM-4 microcomputer, TMS units as DCO, an adapter for communication of the common bus of the SM-4 minicomputer with the IK1 TMS bus developed for their conjugation and devices for simulation of the facility operation is suggested to increase labour productivity of programmers. The SM-4 microcomputer permits to develop programs in the FORTRAN language and to carry out their effective adjustment. A subprogram library for communication with TMS units is compiled, and the technique for testing FORTRAN programs in programmable read-only memory TMS as well as start of these programs are developed

  16. Fuel cycle management in Finland

    International Nuclear Information System (INIS)

    Vaeyrynen, H.; Mikkola, I.

    1987-01-01

    Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

  17. Monitoring fuel consumption for reciprocating engines

    International Nuclear Information System (INIS)

    Zebelean, D.C.

    1991-01-01

    This paper reports that Northwest Pipeline Corp. has experienced substantial fuel cost savings through a program that provides continuous monitoring of predicted consumption against actual usage for reciprocating engines. With the continuous monitoring program, the company can gauge specific fuel consumption, flag inefficient engines and determine possible problems with computer software or hardware used to measure engine operations. The plan was initiated as part of an overall effort to reduce pipe line operating cost to remain market competitive by reducing cost of services to the customer. One of the factors in determining Northwest Pipeline's cost of service is the cost of fueling engines to transport natural gas. In 1990, Northwest consumed approximately 9,600,000 MMBtu in engine fuel. Fuel gas always has been accounted for and measured. However, the thermal efficiency of the fuel gas consumed was never quantified. Engineering management set a goal in 1989 to audit fuel consumption monthly on the basis of actual fuel consumed compared to the manufacturer's predicted consumption curves. The fuel consumption comparison between actual consumption with manufacturer's predicted consumption decreased 4% in six months after the first report was published. Unbalanced engines, faulty spark plugs, fuel valves, engines requiring overhauls, and even computer software and hardware problems were found, based on this report. Total decrease in the comparison of 4% was not all realized. Approximately 1.5% of the decrease was due to revising the manufacturer's fuel consumption curves to correctly predict Northwest's vintage of engine

  18. The use of computed neutron coincidence counting with time interval analysis for the analysis of Fork-measurements on a fresh MOX-LWR fuel assembly under water

    Energy Technology Data Exchange (ETDEWEB)

    Baeten, P.; Bruggeman, M.; Carchon, R

    1998-06-01

    The objective of this study was to investigate the influence of different important parameters on measurement results for various fork-detectors. Computed Neutron Coincidence Counting (CNCC) with Time Interval Analysis (TIA) was used for this study. The performance of the electronics for the different fork-detectors was studied by investigating the deadtime perturbed zone of the Rossi-alpha distribution in TIA. The measurement revealed anomalies in the performance of the electronics of the IAEA BWR and LANL fork-detector. The IAEA PWR fork-detector functioned well and the deadtime parameter was calculated. The optimal setting for the pre delay was investigated and it was found that a pre delay of 10 micro seconds should be considered as an optimum between excluding from analysis data in the deadtime perturbed zone and keeping a high signal-to-noise ratio. For the shift register electronics used with the fork-detectors, a pre delay of only 4.5 micro seconds was used. The study of the pre delay and the deadtime showed that the calculated triples-rate is strongly dependent on these parameters. An accurate determination of the triple-rate in this type of measurements has proven to be quite difficult and requires proper operation of the electronics, a correct pre delay and an accurate deadtime correction formalism. By varying the boron concentration in water, the change of the decay time of the Rossi-alpha distribution was clearly observed. This change is due to the variation of the thermal multiplication. The variation of this decay time with the boron concentration proves that Boehnel's model for fast neutron multiplication is not valid under these measurement conditions and that a model for fast and thermal multiplication should be used in order to obtain unbiased measurement results. CNCC with TIA has proved to be a valuable tool in which parameter settings can be varied a posterori and the optimal setting can be determined for each measurement. Moreover, the

  19. The use of computed neutron coincidence counting with time interval analysis for the analysis of Fork-measurements on a fresh MOX-LWR fuel assembly under water

    International Nuclear Information System (INIS)

    Baeten, P.; Bruggeman, M.; Carchon, R.

    1998-06-01

    The objective of this study was to investigate the influence of different important parameters on measurement results for various fork-detectors. Computed Neutron Coincidence Counting (CNCC) with Time Interval Analysis (TIA) was used for this study. The performance of the electronics for the different fork-detectors was studied by investigating the deadtime perturbed zone of the Rossi-alpha distribution in TIA. The measurement revealed anomalies in the performance of the electronics of the IAEA BWR and LANL fork-detector. The IAEA PWR fork-detector functioned well and the deadtime parameter was calculated. The optimal setting for the pre delay was investigated and it was found that a pre delay of 10 micro seconds should be considered as an optimum between excluding from analysis data in the deadtime perturbed zone and keeping a high signal-to-noise ratio. For the shift register electronics used with the fork-detectors, a pre delay of only 4.5 micro seconds was used. The study of the pre delay and the deadtime showed that the calculated triples-rate is strongly dependent on these parameters. An accurate determination of the triple-rate in this type of measurements has proven to be quite difficult and requires proper operation of the electronics, a correct pre delay and an accurate deadtime correction formalism. By varying the boron concentration in water, the change of the decay time of the Rossi-alpha distribution was clearly observed. This change is due to the variation of the thermal multiplication. The variation of this decay time with the boron concentration proves that Boehnel's model for fast neutron multiplication is not valid under these measurement conditions and that a model for fast and thermal multiplication should be used in order to obtain unbiased measurement results. CNCC with TIA has proved to be a valuable tool in which parameter settings can be varied a posterori and the optimal setting can be determined for each measurement. Moreover, the

  20. 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.)

  1. 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.)

  2. 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.)

  3. 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.)

  4. 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)

  5. Likelihood estimations of vegetative alteration near known or proposed sources of air pollution. 1st annual progress report. [Computer simulation of effects on plants of SO/sub 2/ from fossil-fuel power plants sited in rural areas

    Energy Technology Data Exchange (ETDEWEB)

    Davis, D.D.; Pennypacker, S.P.

    1978-08-01

    There is an increasing trend to locate fossil-fuel electric generating stations in rural areas. These stations emit large quantities of air pollutants capable of causing serious alterations to the surrounding environment. The major phytotoxic air pollutant emitted is sulfur dioxide (SO/sub 2/). The best way to predict the potential impact of new or expanded sources in rural areas would be to develop a computer model to simulate and predict SO/sub 2/ injury to vegetation. The proposed simulator would provide valuable input information for tasks such as site selection and compilation of environmental impact statements. Such a model would also provide data to management operators for the regulation of emissions. The model would also aid in our basic understanding of the complex interactions which influence plant susceptibility to air pollution. Input to the model would consist of biological and physical data and the output would include the probability of SO/sub 2/ injuring vegetation near existing or proposed sites. The model would be validated using a field situation. This report presents the results obtained during the first year of the project (August 1, 1977 to July 31, 1978).

  6. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  7. Uncertainty Analyses of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    Miller, Laurence F.; Preston, J.; Sweder, G.; Anderson, T.; Janson, S.; Humberstone, M.; MConn, J.; Clark, J.

    2008-01-01

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development

  8. ADELA - user interface for fuel charge design

    International Nuclear Information System (INIS)

    Havluj, Frantisek

    2010-01-01

    ADELA is a supporting computer code - ANDREA code add-on - for fuel batch designing and optimization. It facilitates fuel batch planning, evaluation and archival by using graphical user interface. ADELA simplifies and automates the design process and is closely linked to the QUADRIGA system for data library creation. (author)

  9. Uncertainty Analyses of Advanced Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Laurence F. Miller; J. Preston; G. Sweder; T. Anderson; S. Janson; M. Humberstone; J. MConn; J. Clark

    2008-12-12

    The Department of Energy is developing technology, experimental protocols, computational methods, systems analysis software, and many other capabilities in order to advance the nuclear power infrastructure through the Advanced Fuel Cycle Initiative (AFDI). Our project, is intended to facilitate will-informed decision making for the selection of fuel cycle options and facilities for development.

  10. Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

    International Nuclear Information System (INIS)

    Bailescu, A.; Barbu, A.; Din, F.; Dinuta, G.; Dumitru, I.; Musetoiu, A.; Serban, G.; Tomescu, A.

    2013-01-01

    The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

  11. 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

  12. The lumped parameter model for fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W S [Ontario Hydro, Toronto, ON (Canada)

    1996-12-31

    The use of a lumped fuel-pin model in a thermal-hydraulic code is advantageous because of computational simplicity and efficiency. The model uses an averaging approach over the fuel cross section and makes some simplifying assumptions to describe the transient equations for the averaged fuel, fuel centerline and sheath temperatures. It is shown that by introducing a factor in the effective fuel conductivity, the analytical solution of the mean fuel temperature can be modified to simulate the effects of the flux depression in the heat generation rate and the variation in fuel thermal conductivity. The simplified analytical method used in the transient equation is presented. The accuracy of the lumped parameter model has been compared with the results from the finite difference method. (author). 4 refs., 2 tabs., 4 figs.

  13. Thermochemistry of nuclear fuels in advanced reactors

    International Nuclear Information System (INIS)

    Agarwal, Renu

    2015-01-01

    The presence of a large number of elements, accompanied with steep temperature gradient results in dynamic chemistry during nuclear fuel burn-up. Understanding this chemistry is very important for efficient and safe usage of nuclear fuels. The radioactive nature of these fuels puts lot of constraint on regulatory bodies to ensure their accident free operation in the reactors. One of the common aims of advanced fuels is to achieve high burn-up. As burn-up of the fuel increases, chemistry of fission-products becomes increasingly more important. To understand different phenomenon taking place in-pile, many out of-pile experiments are carried out. Extensive studies of thermodynamic properties, phase analysis, thermophysical property evaluation, fuel-fission product clad compatibility are carried out with relevant compounds and simulated fuels (SIMFUEL). All these data are compiled and jointly evaluated using different computational methods to predict fuel behaviour during burn-up. Only when this combined experimental and theoretical information confirms safe operation of the pin, a test pin is prepared and burnt in a test reactor. Every fuel has a different chemistry and different constraints associated with it. In this talk, various thermo-chemical aspects of some of the advanced fuels, mixed carbide, mixed nitride, 'Pu' rich MOX, 'Th' based AHWR fuels and metallic fuels will be discussed. (author)

  14. Substantial Improvements of Fuel Economy

    DEFF Research Database (Denmark)

    Jørgensen, Kaj; Nielsen, Lars H.

    1996-01-01

    The paper evaluates the scope for improving the energy and environmental impacts of road transport by means of electrical and hybrid propulsion. These technologies promise considerable improvements of the fuel economy compared to equivalent vehicles mas well as beneficial effects for the energy...... and traffic systems. A case study concerning passenger cars is analysed by means of computer simulation....

  15. 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.

  16. Radioactive decay properties of CANDU fuel. Volume 1: the natural uranium fuel cycle

    International Nuclear Information System (INIS)

    Clegg, L.J.; Coady, J.R.

    1977-01-01

    The two books of Volume 1 comprise the first in a three-volume series of compilations on the radioactive decay propertis of CANDU fuel and deal with the natural uranium fuel cycle. Succeeding volumes will deal with fuel cycles based on plutonium recycle and thorium. In Volume 1 which is divided into three parts, the computer code CANIGEN was used to obtain the mass, activity, decay heat and toxicity of CANDU fuel and its component isotopes. Data are also presented on gamma spectra and neutron emissions. Part 3 contains the data relating to the plutonium product and the high level wastes produced during fuel reprocessing. (author)

  17. 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

  18. 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

  19. 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.

  20. 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.)

  1. 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.)

  2. 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.)

  3. 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...

  4. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  5. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  6. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  7. Fuel pellet fracture and relocation

    International Nuclear Information System (INIS)

    Walton, L.A.; Husser, D.L.

    1983-01-01

    The model used to describe fuel pellet fracture and relocation is an important feature of a fuel performance computer code. This model becomes especially important if the computer code is principally to be used for the evaluation of pellet clad interaction. The fracture and relocation model being developed for the B and W fuel performance code FUMAC was derived from an extensive data base. Cross sections of irradiated fuel rods were photographically magnified and measured to determine the configuration of the fragments of the fractured fuel pellets. Data, representing a wide range of LWR fuel designs and as-manufactured mechanical configurations, were catalogued and systematically reduced and then correlated as a function of the likely independent variables. These correlations define the key phenomenological behavior patterns which the relocation model must duplicate and indicate which mechanistic approaches are viable explanations of this behavior. The data base covers the burnup range from approximately one to 35 GWd/mtU and linear heat rates from less than 100 to nearly 700 W/Cm. This paper presents the correlated data base and the methods used to derive and interpret it. It was determined from this data base that pellet cracking is initially both power level and burnup dependent but tends to saturate eventually with continued steady irradiation. Fuel pellet relocation was found to be much more extensive than would be deduced from thermal considerations alone. Even at very low burnups fuel fragments were found to move outward until restrained by the cladding. The results also suggest that changes in internal resistance to heat flow within the pellets due to the opening of cracks may be as important as peripheral gap changes to the thermal modeler. The transient response and thermal implications of this model are recommended as primary areas for future investigation

  8. Validation of computer codes and modelling methods for giving proof of nuclear safety of transport and storage of spent VVER-type nuclear fuels. Pt. 2. Criticality safety during transport and storage of spent VVER fuel elements. Final report; Einschaetzung von Rechenprogrammen und Methoden zum Nachweis der nuklearen Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoffen. T. 2. Kritikalitaetssicherheit bei Transport und Lagerung von WWER-Brennelementen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-03-15

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k{sub eff} is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [Deutsch] Der Bericht enthaelt die Ergebnisse von Untersuchungen zur Validierung von Rechenprogrammen, welche zum Nachweis der nukelaren Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoff der KKW Greifswald und Rheinsberg eingesetzt wurden. Es werden eine Charakteristik des abgebrannten Brennstoffs (Nuklidkonzentrationen, Neutronenquellstaerke, Gammaspektrum, Nachzerfallsleistung) - berechnet mit verschiedenen Programmen - und Ortsdosisleistungen (z.B. in der Umgebung eines Transportbehaelters) - basierend auf den verschiedenen Quelltermen - angegeben. Differenzen und Ursachen werden diskutiert. Die Ergebnisse zeigen, dass trotz der Differenzen in den Quelltermen alle strahlenschutztechnisch relevanten Aussagen unbeeinflusst bleiben. Fuer die Einschaetzung des Gueltigkeitsbereiches des Monte-Carlo-Programms OMEGA wurden ca. 200 kritische Experimente mit LWR-Brennstoff unter besonderer Beruecksichtigung

  9. 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)

  10. Fuel channel refilling

    International Nuclear Information System (INIS)

    Shoukri, M.; Abdul-Razzak, A.

    1992-11-01

    This report extends the work reported in document INFO-0370 on fuel channel refilling by providing analysis of the refilling tests conducted using the RD-14 and RD-14M test facilities. The analysis focuses on the general thermal-hydraulic characteristics of the facilities following various small and large inlet and outlet header breaks with emergency coolant injection. The two-fluid model thermal-hydraulic computer code CATHENA was tested against results obtained from selected experiments carried out in the two facilities. Conclusions related to the effect of break size, mode of emergency core injection, primary pump operation and parallel channels are presented. (Author) (116 figs., 17 tabs., 53 refs.)

  11. Fuel cycle kinetics

    International Nuclear Information System (INIS)

    Maudlin, P.J.

    1979-01-01

    A theoretical methodology describing the time dependent growth of large populations of nuclear power reactors of different types is pursued. This methodology is based on the apparent close analogy between the time dependent variations of neutrons and of fuel in nuclear reactors. Methods for the realistic projection of reactor populations, as they develop in a reactor park, are provided using the point park model as kernel in a superposition of reactor deployment elements that form a realistic park scenario. Typical deployment strategy results are presented illustrating the theoretical and computational advantages of the point park model methodology

  12. 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.)

  13. Used fuel disposition in crystalline rocks

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Y. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jerden, James L. [Argonne National Lab. (ANL), Argonne, IL (United States); Copple, Jacqueline M. [Argonne National Lab. (ANL), Argonne, IL (United States); Cruse, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Ebert, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Buck, E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eittman, R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tinnacher, R. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Tournassat, Christophe. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Davis, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Viswanathan, H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dittrich, T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hyman, F. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karra, S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Makedonska, N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reimus, P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Joseph, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-01

    The U.S. Department of Energy Office of Nuclear Energy, Office of Fuel Cycle Technology established the Used Fuel Disposition Campaign (UFDC) in fiscal year 2010 (FY10) to conduct the research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel and high level nuclear waste. The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media.

  14. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  15. Seismic analysis of freestanding fuel racks

    International Nuclear Information System (INIS)

    Gilmore, C.B.

    1982-01-01

    This paper presents a nonlinear transient dynamic time-history analysis of freestanding spent fuel storage racks subjected to seismic excitation. This type of storage rack is structurally unrestrained and submerged in water in the spent fuel pool of a nuclear power complex, holds (spent) fuel assemblies which have been removed from the reactor core. Nonlinearities in the fuel rack system include impact between the fuel assembly and surrounding cell due to clearances between them, friction due to sliding between the fuel rack support structure and spent fuel pool floor, and the lift-off of the fuel rack support structure from the spent fuel pool floor. The analysis of the fuel rack system includes impacting due to gap closures, energy losses due to impacting bodies, Coulomb damping between sliding surfaces, and hydrodynamic mass effects. Acceleration time history excitation development is discussed. Modeling considerations, such as the initial status of nonlinear elements, number of mode shapes to include in the analysis, modal damping, and integration time-step size are presented. The response of the fuel rack subjected to two-dimensional seismic excitation is analyzed by the modal superposition method, which has resulted in significant computer cost savings when compared to that of direct integration

  16. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Bolme, A.B.

    1996-01-01

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  17. Simulation of leaking fuel rods

    International Nuclear Information System (INIS)

    Hozer, Z.

    2006-01-01

    The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of 123 I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the 123 I release from failed fuel rods during transients

  18. Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440. Preliminary assessment of operating efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gagarinskiy, Alexey [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2017-09-15

    Since the introduction of VVERs-440, their fuel assemblies are subject to ongoing improvements. Until now, the basic structural parameters of fuel, such as rod diameter of 9.1 mm, have never changed. This paper focuses on computational estimates of basic neutronic parameters of the fuel cycle that involves assemblies consisting of fuel rods with diameter reduced to 8.9 mm.

  19. 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

  20. 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.)

  1. 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.)

  2. 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. )

  3. 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.)

  4. 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.)

  5. Spent fuel reprocessing system availability definition by process simulation

    International Nuclear Information System (INIS)

    Holder, N.; Haldy, B.B.; Jonzen, M.

    1978-05-01

    To examine nuclear fuel reprocessing plant operating parameters such as maintainability, reliability, availability, equipment redundancy, and surge storage requirements and their effect on plant throughput, a computer simulation model of integrated HTGR fuel reprocessing plant operations is being developed at General Atomic Company (GA). The simulation methodology and the status of the computer programming completed on reprocessing head end systems is reported

  6. Solid TRU fuels and fuel cycle technology

    International Nuclear Information System (INIS)

    Ogawa, Toru; Suzuki, Yasufumi

    1997-01-01

    Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

  7. Used fuel packing plant for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

    2015-07-01

    Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

  8. How well does ORIGEN predict spent LWR fuel characteristics

    International Nuclear Information System (INIS)

    Mailen, J.C.; Roddy, J.W.

    1987-01-01

    The ORIGEN computer code is widely used to estimate the radionuclide content (actinides, activation and fission products) of irradiated reactor fuel and the resultant heat generation and radiation levels associated with such fuel. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of the fuel. This survey summarizes the fuel data available in the open literature and, where given, the calculated values by ORIGEN. Plans for additional analyses of well-characterized reactor fuel samples to improve the validation of ORIGEN2 are discussed

  9. Automated Fuel Element Closure Welding System

    International Nuclear Information System (INIS)

    Wahlquist, D.R.

    1993-01-01

    The Automated Fuel Element Closure Welding System is a robotic device that will load and weld top end plugs onto nuclear fuel elements in a highly radioactive and inert gas environment. The system was developed at Argonne National Laboratory-West as part of the Fuel Cycle Demonstration. The welding system performs four main functions, it (1) injects a small amount of a xenon/krypton gas mixture into specific fuel elements, and (2) loads tiny end plugs into the tops of fuel element jackets, and (3) welds the end plugs to the element jackets, and (4) performs a dimensional inspection of the pre- and post-welded fuel elements. The system components are modular to facilitate remote replacement of failed parts. The entire system can be operated remotely in manual, semi-automatic, or fully automatic modes using a computer control system. The welding system is currently undergoing software testing and functional checkout

  10. Fuel analysis code FAIR and its high burnup modelling capabilities

    International Nuclear Information System (INIS)

    Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs

  11. 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

  12. Computer generated movies at Hanford

    International Nuclear Information System (INIS)

    Lewis, C.H.; Fox, G.L.

    1979-10-01

    The message contained in the results of a large computer program is often difficult to present to large groups of people. This difficulty may be overcome by using 16mm color movie techniques. This presentation shows the results of directly using computer output to explain a story about fuel behavior during a power transient

  13. Analysis of irradiation temperature in fuel rods of OGL-1 fuel assembly

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Minato, Kazuo; Ikawa, Katsuichi; Iwamoto, Kazumi

    1984-10-01

    Irradiation temperature in the fuel rods of 5th OGL-1 fuel assembly was analysed by the system composed by STPDSP2 and TRUMP codes. As the measured input-data, following parameters were allowed for; circumferential heating distribution around the fuel rod, which was measured in the JMTR critical assembly, axial heating distribution through the fuel rod, ratio of peak heatings of three fuel rods, and pre- and post-irradiation outer radii of the fuel compacts and inner radii of the graphite sleeves, which had been measured in PIE of the 5th OGL-1 fuel assembly. In computation the axial distributions of helium coolant temperature through the fuel rod and the heating value of each fuel rod were, firstly, calculated as input data for TRUMP. The TRUMP calculation yielded the temperatures which were fitted in those measured by all of the thermo-couples installed in the fuel rods, by adjusting only the value of the surface heat transfer coefficient, and consequently, the temperatures in all portions of the fuel rod were obtained. The apparent heat transfer coefficient changed to 60% of the initial values in the middle period of irradiation. For this reduction it was deduced that shoot had covered the surface of the fuel rod during irradiation, which was confirmed in PIE. Beside it, several things were found in this analysis. (author)

  14. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  15. Fuel element loading system

    International Nuclear Information System (INIS)

    Arya, S.P; s.

    1978-01-01

    A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

  16. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  17. CFD thermal-hydraulic analysis of a CANDU fuel channel

    International Nuclear Information System (INIS)

    Catana, A.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2009-01-01

    This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational fluid dynamics) methodology approach. Limited computer power available at Bucharest University POLITEHNICA forced the authors to analyse only segments of fuel channel namely the significant ones: fuel bundle junctions with adjacent segments, fuel bundle spacer planes with adjacent segments, regular segments of fuel bundles. The computer code used is FLUENT. Fuel bundles contained in pressure tubes forms a complex flow domain. The flow is characterized by high turbulence and in some parts of fuel channel also by multi-phase flow. The flow in the fuel channel has been simulated by solving the equations for conservation of mass and momentum. For turbulence modelling the standard k-e model is employed although other turbulence models can be used as well. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Since we consider only some relatively short segments of a CANDU fuel channel we can assume, for this starting stage, that heat transfer is not very important for these short segments of fuel channel. The boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. In this paper we present results for Standard CANDU 6 Fuel Bundles as a basis for CFD thermal-hydraulic analysis of INR proposed SEU43 and other new nuclear fuels. (authors)

  18. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  19. Artificial fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hamon, L L.W.

    1918-08-20

    Lignite, peat, sud, leaf-mold, or shale, or two or more of these raw carbonaceous materials are mixed with cellulose material, such as sawdust, silica, alkali, and tar or pitch, or residues from tar or pitch, or residues from the distillation of oils, and the mixture is molded into blocks. Other carbonaceous materials, such as graphite, anthracite, or coal-dust, coke, breeze, or culm, and mineral substances, such as iron and manganese ores, may be added. A smokeless fuel can be obtained by coking the blocks in the usual way in retorts.

  20. MTR radiological database for SRS spent nuclear fuel facilities

    International Nuclear Information System (INIS)

    Blanchard, A.

    2000-01-01

    A database for radiological characterization of incoming Material Test Reactor (MTR) fuel has been developed for application to the Receiving Basin for Offsite Fuels (RBOF) and L-Basin spent fuel storage facilities at the Savannah River Site (SRS). This database provides a quick quantitative check to determine if SRS bound spent fuel is radiologically bounded by the Reference Fuel Assembly used in the L-Basin and RBOF authorization bases. The developed database considers pertinent characteristics of domestic and foreign research reactor fuel including exposure, fuel enrichment, irradiation time, cooling time, and fuel-to-moderator ratio. The supplied tables replace the time-consuming studies associated with authorization of SRS bound spent fuel with simple hand calculations. Additionally, the comprehensive database provides the means to overcome resource limitations, since a series of simple, yet conservative, hand calculations can now be performed in a timely manner and replace computational and technical staff requirements

  1. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  2. Hydrogen and fuel cell research: Institute for Integrated Energy Systems (IESVic)

    International Nuclear Information System (INIS)

    Pitt, L.

    2006-01-01

    Vision: IESVic's mission is to chart feasible paths to sustainable energy. Current research areas of investigation: 1. Energy system analysis 2. Computational fuel cell engineering; Fuel cell parameter measurement; Microscale fuel cells 3. Hydrogen dispersion studies for safety codes 4. Active magnetic refrigeration for hydrogen liquifaction and heat transfer in metal hydrides 5. Hydrogen and fuel cell system integration (author)

  3. An introduction to current modeling techniques in nuclear fuel performance analysis

    International Nuclear Information System (INIS)

    Faya, S.C.S.

    1981-10-01

    An overview of several important phenomena considered in fuel performance analysis, such as fuel restructuring, fuel swelling, fission gas release, pellet cracking, etc... are provided. All these phenomena are independent. For the purpose of illustration of integral fuel rod analysis the LIFE-1 computer code is discussed. (Author) [pt

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Hiraiwa, Koji; Ueda, Makoto

    1989-01-01

    In a fuel assembly used for a light water cooled reactor such as a BWR type reactor, a water rod is divided axially into an upper outer tube and a lower outer tube by means of a plug disposed from the lower end of a water rod to a position 1/4 - 1/2 of the entire length for the water rod. Inlet apertures and exit apertures for moderators are respectively perforated for the divided outer tube and upper and lower portions. Further, an upper inner tube with less neutron irradiation growing amount than the outer tube is perforated on the plug in the outer tube, while a lower inner tube with greater neutron irradiation growing amount than the outer tube is suspended from the lower surface of the plug in the outer tube. Then, the opening area for the exit apertures disposed to the upper outer tube and the lower outer tube is controlled depending on the difference of the neutron irradiation growing amount between the upper inner tube and the upper outer tube, and the difference of the neutron irradiation growing amount between the lower inner tube and the lower outer tube. This enables effective spectral shift operation and improve the fuel economy. (T.M.)

  5. Fuel Burn Estimation Model

    Science.gov (United States)

    Chatterji, Gano

    2011-01-01

    Conclusions: Validated the fuel estimation procedure using flight test data. A good fuel model can be created if weight and fuel data are available. Error in assumed takeoff weight results in similar amount of error in the fuel estimate. Fuel estimation error bounds can be determined.

  6. Effects of cooling time on a closed LWR fuel cycle

    International Nuclear Information System (INIS)

    Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

    2012-01-01

    In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

  7. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  8. 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.

  9. Design of fuel loading for Bohunice V-1 Unit 2 reaktor for fuel cycle No.19

    International Nuclear Information System (INIS)

    Majercik, J.

    1998-01-01

    The report contains description of the design of fuel loading for the fuel cycle No. 19 in the V-1 Bohunice Unit 2 reactor. Input data and computer codes used for the development of the design are shown. The fuel loading is characterized by the assortment of the fuel loaded and by the scheme of re shuffling of assemblies in the core. An evaluation of basic neutronic core parameters as relates to the compliance with safety criteria is a part of the report as well

  10. KMRR fuel design

    International Nuclear Information System (INIS)

    Son, D.S.; Sim, B.S.; Kim, T.R.; Hwang, W.; Kim, B.G.; Ku, Y.H.; Lee, C.B.; Lim, I.C.

    1992-06-01

    KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

  11. Fuel Property Blend Model

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wagnon, Scott J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zhang, Kuiwen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kukkadapu, Goutham [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-01-12

    The object of this project is to develop chemical models and associated correlations to predict the blending behavior of bio-derived fuels when mixed with conventional fuels like gasoline and diesel fuels.

  12. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

  13. Fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    1980-01-01

    Apparatus is described for loading a predetermined amount of nuclear fuel pellets into nuclear fuel elements and particularly for the automatic loading of fuel pellets from within a sealed compartment. (author)

  14. Potential health hazard of nuclear fuel waste and uranium ore

    International Nuclear Information System (INIS)

    Mehta, K.; Sherman, G.R.; King, S.G.

    1991-06-01

    The variation of the radioactivity of nuclear fuel waste (used fuel and fuel reprocessing waste) with time, and the potential health hazard (or inherent radiotoxicity) resulting from its ingestion are estimated for CANDU (Canada Deuterium Uranium) natural-uranium reactors. Four groups of radionuclides in the nuclear fuel waste are considered: actinides, fission products, activation products of zircaloy, and activation products of fuel impurities. Contributions from each of these groups to the radioactivity and to the potential health hazard are compared and discussed. The potential health hazard resulting from used fuel is then compared with that of uranium ore, mine tailings and refined uranium (fresh fuel) on the basis of equivalent amounts of uranium. The computer code HAZARD, specifically developed for these computations, is described

  15. Stationary Liquid Fuel Fast Reactor

    International Nuclear Information System (INIS)

    Yang, Won Sik; Grandy, Andrew; Boroski, Andrew; Krajtl, Lubomir; Johnson, Terry

    2015-01-01

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  16. Stationary Liquid Fuel Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Grandy, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Boroski, Andrew [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, Lubomir [Argonne National Lab. (ANL), Argonne, IL (United States); Johnson, Terry [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-09-30

    For effective burning of hazardous transuranic (TRU) elements of used nuclear fuel, a transformational advanced reactor concept named SLFFR (Stationary Liquid Fuel Fast Reactor) was proposed based on stationary molten metallic fuel. The fuel enters the reactor vessel in a solid form, and then it is heated to molten temperature in a small melting heater. The fuel is contained within a closed, thick container with penetrating coolant channels, and thus it is not mixed with coolant nor flow through the primary heat transfer circuit. The makeup fuel is semi- continuously added to the system, and thus a very small excess reactivity is required. Gaseous fission products are also removed continuously, and a fraction of the fuel is periodically drawn off from the fuel container to a processing facility where non-gaseous mixed fission products and other impurities are removed and then the cleaned fuel is recycled into the fuel container. A reference core design and a preliminary plant system design of a 1000 MWt TRU- burning SLFFR concept were developed using TRU-Ce-Co fuel, Ta-10W fuel container, and sodium coolant. Conservative design approaches were adopted to stay within the current material performance database. Detailed neutronics and thermal-fluidic analyses were performed to develop a reference core design. Region-dependent 33-group cross sections were generated based on the ENDF/B-VII.0 data using the MC2-3 code. Core and fuel cycle analyses were performed in theta-r-z geometries using the DIF3D and REBUS-3 codes. Reactivity coefficients and kinetics parameters were calculated using the VARI3D perturbation theory code. Thermo-fluidic analyses were performed using the ANSYS FLUENT computational fluid dynamics (CFD) code. Figure 0.1 shows a schematic radial layout of the reference 1000 MWt SLFFR core, and Table 0.1 summarizes the main design parameters of SLFFR-1000 loop plant. The fuel container is a 2.5 cm thick cylinder with an inner radius of 87.5 cm. The fuel

  17. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  18. Characteristics of spent nuclear fuel

    International Nuclear Information System (INIS)

    Notz, K.J.

    1988-04-01

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs

  19. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  20. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  1. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  2. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2001-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  3. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2000-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  4. CANDU-PHW fuel management

    International Nuclear Information System (INIS)

    Frescura, G.M.; Wight, A.L.

    1982-01-01

    This report covers the material presented in a series of six lectures at the Winter College on Nuclear Physics and Reactors held at the International Centre for Theoretical Physics in Trieste, Italy, Jan 22 - March 28, 1980. The report deals with fuel management in natural uranium fuelled CANDU-PHW reactors. Assuming that the reader has a basic knowledge of CANDU core physics, some of the reactor systems which are more closely related to fuelling are described. This is followed by a discussion of the methods used to calculate the power distribution and perform fuel management analyses for the equilibrium core. A brief description of some computer codes used in fuel management is given, together with an overview of the calculations required to provide parameters for core design and support the accident analysis. Fuel scheduling during approach to equilibrium and equilibrium is discussed. Fuel management during actual reactor operation is discussed with a review of the operating experience for some of the Ontario Hydro CANDU reactors. (author)

  5. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL)Established to investigate, integrate, testand verifyperformance and technology readiness offuel cell systems and fuel reformers for use with...

  6. Fuel performance experience

    International Nuclear Information System (INIS)

    Sofer, G.A.

    1986-01-01

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  7. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  8. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

    Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

    1976-01-01

    The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

  9. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  10. Measurement of the fuel temperature and the fuel-to-coolant heat transfer coefficient of Super Phenix 1 fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1995-12-01

    A new measurement method for measuring the mean fuel temperature as well as the fuel-to-coolant heat transfer coefficient of fast breeder reactor subassemblies (SA) is reported. The method is based on the individual heat balance of fuel SA's after fast reactor shut-downs and uses only the plants normal SA outlet temperature and neutron power signals. The method was used successfully at the french breeder prototype Super Phenix 1. The mean SA fuel temperature as well as the heat transfer coefficient of all SPX SA's have been determined at power levels between 15 and 90% of nominal power and increasing fuel burn-up from 3 to 83 EFPD (Equivalent of Full Power-Days). The measurements also provided fuel and whole SA time constants. The estimated accuracy of measured fuel parameters is in the order of 10%. Fuel temperatures and SA outlet temperature transients were also calculated with the SPX1 systems code DYN2 for exactly the same fuel and reactor operating parameters as in the experiments. Measured fuel temperatures were higher than calculated ones in all cases. The difference between measured and calculated core mean values increases from 50 K at low power to 180 K at 90% n.p. This is about the double of the experimental error margins. Measured SA heat transfer coefficients are by nearly 20% lower than corresponding heat transfer parameters used in the calculations. Discrepancies found between measured and calculated results also indicate that either the transient heat transfer in the gap between fuel and cladding (gap conductance) might not be exactly reproduced in the computer code or that the gap in the fresh fuel was larger than assumed in the calculations. (orig.) [de

  11. Studies on validation possibilities for computational codes for criticality and burnup calculations of boiling water reactor fuel; Untersuchungen zu Validierungsmoeglichkeiten von Rechencodes fuer Kritikalitaets- und Abbrandrechnungen von Siedewasserreaktor-Brennstoff

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik

    2017-06-15

    The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.

  12. Computer Refurbishment

    International Nuclear Information System (INIS)

    Ichiyen, Norman; Chan, Dominic; Thompson, Paul

    2004-01-01

    The major activity for the 18-month refurbishment outage at the Point Lepreau Generating Station is the replacement of all 380 fuel channel and calandria tube assemblies and the lower portion of connecting feeder pipes. New Brunswick Power would also take advantage of this outage to conduct a number of repairs, replacements, inspections and upgrades (such as rewinding or replacing the generator, replacement of shutdown system trip computers, replacement of certain valves and expansion joints, inspection of systems not normally accessible, etc.). This would allow for an additional 25 to 30 years. Among the systems to be replaced are the PDC's for both shutdown systems. Assessments have been completed for both the SDS1 and SDS2 PDC's, and it has been decided to replace the SDS2 PDCs with the same hardware and software approach that has been used successfully for the Wolsong 2, 3, and 4 and the Qinshan 1 and 2 SDS2 PDCs. For SDS1, it has been decided to use the same software development methodology that was used successfully for the Wolsong and Qinshan called the I A and to use a new hardware platform in order to ensure successful operation for the 25-30 year station operating life. The selected supplier is Triconex, which uses a triple modular redundant architecture that will enhance the robustness/fault tolerance of the design with respect to equipment failures

  13. Low stoichiometry operation of a proton exchange membrane fuel cell employing the interdigitated flow field

    DEFF Research Database (Denmark)

    Berning, Torsten; Kær, Søren Knudsen

    2012-01-01

    A multiphase fuel cell model based on computational fluid dynamics is used to investigate the possibility of operating a proton exchange membrane fuel cell at low stoichiometric flow ratios (ξ gases. A case study...

  14. Application of the Advanced Distillation Curve Method to Fuels for Advanced Combustion Engine Gasolines

    KAUST Repository

    Burger, Jessica L.; Schneider, Nico; Bruno, Thomas J.

    2015-01-01

    . To make changes in the most time- and cost-effective manner, it is imperative that new computational tools and surrogate fuels are developed. Currently, sets of fuels are being characterized by industry groups, such as the Coordinating Research Council

  15. From Neurons to Brainpower: Cognitive Neuroscience and Brain-Based Learning

    Science.gov (United States)

    Phillips, Janet M.

    2005-01-01

    We have learned more about the brain in the past five years than the previous 100. Neuroimaging, lesion studies, and animal studies have revealed the intricate inner workings of the brain and learning. Synaptogenesis, pruning, sensitive periods, and plasticity have all become accepted concepts of cognitive neuroscience that are now being applied…

  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. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  18. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  19. Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel

    International Nuclear Information System (INIS)

    Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong

    2015-01-01

    The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)

  20. An equipment for the dimensional characterization of irradiated fuel channels

    International Nuclear Information System (INIS)

    Cederquist, H.

    1985-01-01

    The reuse of irradiated fuel channels in BWRs is highly beneficial. However, one prerequisite for reuse of a fuel channel is the detailed knowledge of its dimensions, which are affected by irradiation and pressure drop during operation. Therefore an equipment for fast and accurate dimensional measurement of irradiated fuel channels has been developed. The measurements are carried out when the fuel assembly is supported in the same manner as in the reactor core. The equipment utilizes stationary ultrasonic transducers that measure the fuel channel at a number of predetermined axial levels. Measurement data are fed into a computer which calculates the requested dimensional characteristics such as transversal flatness, bow, twist, side perpendicularity etc. Data are automatically printed for subsequent evaluation. Measurements can be performed both when the fuel channel is placed on a fuel bundle and on an empty fuel channel