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Sample records for burnup code excel

  1. A VVER-1000 LEU and MOX assembly computational benchmark analysis using the lattice burnup code EXCEL

    Energy Technology Data Exchange (ETDEWEB)

    Thilagam, L. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: thilagam@igcar.gov.in; Sunil Sunny, C. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India); Jagannathan, V. [Light Water Reactor Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: v_jagan1952@rediffmail.com; Subbaiah, K.V. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2009-05-01

    Utilization of Mixed Uranium-Plutonium Oxide (MOX) fuel in VVER-1000 reactors envisages the core physics analysis using computational methods and validation of the related computer codes. Towards this objective, an international experts group has been established at OECD/NEA. The experts group facilitates sharing of existing information on physics parameters and fuel behaviour. Several benchmark exercises have been proposed by them with intent to investigate the core physics behaviour of a VVER-1000 reactor loaded with 2/3rd of low enriched uranium (LEU) fuel assemblies (FA) and 1/3rd of weapons grade mixed oxide (MOX) FA. In the present study an attempt is made to analyse 'AVVER-1000LEUandMOXAssemblyComputationalBenchmark' and predict the neutronics behaviour at the lattice level. The lattice burnup code EXCEL, developed at Light Water Reactor Physics Section, BARC is employed for this task. The EXCEL code uses the 172 energy group 'JEFF31GX' cross-section library in WIMS-D format. Assembly level fuel depletion calculations are performed up to a burnup of 40 MWD/kg of heavy metal (HM). Studies are made for the parametric variations of fuel and moderator temperatures, coolant density and boron content in the coolant. Both operational and off-normal states are analysed to determine the corresponding infinite neutron multiplication factor (k{sub {infinity}}). Pin wise isotopic compositions are computed as a function of burnup. Isotopic compositions in different annular regions of Uranium-Gadolinium (UGD) pin, fission rate distributions in UGD, UO{sub 2} and MOX pin cells are also computed. The predicted results are compared with the benchmark mean results.

  2. Burnup calculations using serpent code in accelerator driven thorium reactors

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    Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.

    2013-07-15

    In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)

  3. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

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    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  4. Thermal behavior analysis of PWR fuel during RIA at various fuel burnups using modified theatre code

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    Nawaz Amjad

    2016-01-01

    Full Text Available The fuel irradiation and burnup causes geometrical and dimensional changes in the fuel rod which affects its thermal resistance and ultimately affects the fuel rod behavior during steady-state and transient conditions. The consistent analysis of fuel rod thermal performance is essential for precise evaluation of reactor safety in operational transients and accidents. In this work, analysis of PWR fuel rod thermal performance is carried out under steady-state and transient conditions at different fuel burnups. The analysis is performed by using thermal hydraulic code, THEATRe. The code is modified by adding burnup dependent fuel rod behavior models. The original code uses as-fabricated fuel rod dimensions during steady-state and transient conditions which can be modified to perform more consistent reactor safety analysis. AP1000 reactor is considered as a reference reactor for this analysis. The effect of burnup on steady-state fuel rod parameters has been investigated. For transient analysis, hypothetical reactivity initiated accident was simulated by considering a triangular power pulse of variable pulse height (relative to the full power reactor operating conditions and pulse width at different fuel burnups which corresponds to fresh fuel, low and medium burnup fuels. The effect of power pulse height, pulse width and fuel burnup on fuel rod temperatures has been investigated. The results of reactivity initiated accident analysis show that the fuel failure mechanisms are different for fresh fuel and fuel at different burnup levels. The fuel failure in fresh fuel is expected due to fuel melting as fuel temperature increases with increase in pulse energy (pulse height. However, at relatively higher burnups, the fuel failure is expected due to cladding failure caused by strong pellet clad mechanical interaction, where, the contact pressure increases beyond the cladding yield strength.

  5. Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.

    Science.gov (United States)

    Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K

    2013-10-01

    Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup. Copyright © 2013 Elsevier Ltd. All rights reserved.

  6. Validation of depletion codes for burnup credit evaluation of LWR assemblies

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    Ranta-aho, A. [Technical Research Centre of Finland VTT, POB 1000, 02044-VTT (Finland)

    2006-07-01

    This paper reports the comparison of the CASMO-4E predictions with the radiochemical assay data from assemblies irradiated in Takahama-3 PWR and Fukushima-Daini-2 BWR, and the most recently reported spent fuel data from the VVER-440 assembly irradiated in Novovoronezh 4. Some of the calculations were repeated with the ABURN burnup code, which is a combination of the MCNP4C Monte Carlo code and the ORIGEN2 depletion code. The cross section libraries applied were based on the ENDF/B-VI and the JEF-2.2 data. (authors)

  7. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  8. A multi-platform linking code for fuel burnup and radiotoxicity analysis

    Science.gov (United States)

    Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.

    2014-02-01

    A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.

  9. Fuel modeling at high burn-up: recent development of the GERMINAL code

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    Melis, J.-C.; Piron, J.-P.; Roche, L.

    1993-09-01

    In the frame of research and development on fast breeder reactors fuels, CEA/DEC is developing the computer code GERMINAL to study fuel pin thermal and mechanical behaviour during steady-state and accidental conditions. The development of the GERMINAL 1 code is foreseen in two steps: (1) The GERMINAL 1-1 version which is presently delivered fully documented with a physical qualification guaranteed up to 8 at%. (2) The GERMINAL 1-2 version which, in addition to what is presently treated in GERMINAL 1-1, includes the treatment of high burn-up effects on the the fission gas release and the fuel-clad interface (called JOG). The validation of GERMINAL 1-2 is presently in progress and will include specific experiments (JOG tests) performed in the CABRI reactor.

  10. Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

    Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

  11. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  12. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

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    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  13. Quantification of the computational accuracy of code systems on the burn-up credit using experimental re-calculations; Quantifizierung der Rechengenauigkeit von Codesystemen zum Abbrandkredit durch Experimentnachrechnungen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik

    2014-06-15

    In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.

  14. Estimation of the Fuel Depletion Code Bias and Uncertainty in Burnup-Credit Criticality Analysis

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    Kim, Jong Woon; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Lee, Sang Jin; Bae, Chang Yeal [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

    2006-07-01

    In the past, criticality safety analyses for commercial light-water-reactor (LWR) spent nuclear fuel (SNF) storage and transportation canisters assumed the spent fuel to be fresh (unirradiated) fuel with uniform isotopic compositions. This fresh-fuel assumption provides a well-defined, bounding approach to the criticality safety analysis that eliminates concerns related to the fuel operating history, and thus considerably simplifies the safety analysis. However, because this assumption ignores the inherent decrease in reactivity as a result of irradiation, it is very conservative. The concept of taking credit for the reduction in reactivity due to fuel burnup is commonly referred to as burnup credit. Implementation of burnup credit requires the computational prediction of the nuclide inventories (compositions) for the dominant fissile and absorbing nuclide species in spent fuel. In addition to that, the bias and uncertainty in the predicted concentration of all nuclides used in the analysis be established by comparisons of calculated and measured radiochemical assay data. In this paper, three methods for considering the bias and uncertainty will be reviewed. The estimated bias and uncertainty that the results of 3rd method are presented.

  15. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

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    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  16. VVER-1000 MOX Core Computational Benchmark analysis using indigenous codes EXCEL, TRIHEX-FA and HEXPIN

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    Thilagam, L. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)], E-mail: thilagam@igcar.gov.in; Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)], E-mail: v_jagan1952@rediffmail.com; Sunil sunny, C.; Subbaiah, K.V. [AERB-Safety Research Institute, Kalpakkam, Tamilnadu 603 102 (India)

    2009-10-15

    Validation studies based on the analysis of theoretical benchmarks play a key role in the identification of deficiencies in the reactor physics design computational codes and the associated nuclear data libraries. Implementation of improvements, if any, in theoretical models and the choice of appropriate nuclear data libraries help in enhancing the accuracy of calculations. As part of the effort for the validation of computer codes for plutonium utilization in VVER type reactors, the indigenous codes EXCEL, TRIHEX-FA and HEXPIN, developed at Light Water Reactor Physics Section (LWRPS), RPDD, BARC, and the associated nuclear data library (JEF22XS), were employed to analyse 'VVER-1000 MOX Core Computational Benchmark'. The few group homogenized parameters of assembly cell or individual lattice cells were obtained by the hexagonal lattice burn-up code EXCEL and the core diffusion calculations were then performed using hexagonal assembly geometric code TRIHEX-FA or the pin-by-pin diffusion code HEXPIN. VVER-1000 reactor core loaded with 2/3rd of Low-Enriched Uranium (LEU) fuel assemblies (FAs) and 1/3rd of weapons grade MOX FAs was investigated. Effective multiplication factors and assembly average fission reaction rate distributions have been calculated for various reactor state descriptions using 3-D diffusion theory codes TRIHEX-FA and HEXPIN. Further, estimate of detailed pin-by-pin fission reaction rate distributions of a few selected assemblies were made for the normal working state of the reactor using pin-by-pin core simulation code HEXPIN. A comparison of results was done with the reported Monte Carlo (MC) values of the benchmark and in most cases good agreement was observed with the benchmark results.

  17. Monte Carlo burnup code acceleration with the correlated sampling method. Preliminary test on an UOX cell with TRIPOLI-4{sup R}

    Energy Technology Data Exchange (ETDEWEB)

    Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, Service d' Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)

    2013-07-01

    For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)

  18. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

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    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  19. Estimating NIRR-1 burn-up and core life time expectancy using the codes WIMS and CITATION

    Science.gov (United States)

    Yahaya, B.; Ahmed, Y. A.; Balogun, G. I.; Agbo, S. A.

    The Nigeria Research Reactor-1 (NIRR-1) is a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria. The reactor went critical with initial core excess reactivity of 3.77 mk. The NIRR-1 cold excess reactivity measured at the time of commissioning was determined to be 4.97 mk, which is more than the licensed range of 3.5-4 mk. Hence some cadmium poison worth -1.2 mk was inserted into one of the inner irradiation sites which act as reactivity regulating device in order to reduce the core excess reactivity to 3.77 mk, which is within recommended licensed range of 3.5 mk and 4.0 mk. In this present study, the burn-up calculations of the NIRR-1 fuel and the estimation of the core life time expectancy after 10 years (the reactor core expected cycle) have been conducted using the codes WIMS and CITATION. The burn-up analyses carried out indicated that the excess reactivity of NIRR-1 follows a linear decreasing trend having 216 Effective Full Power Days (EFPD) operations. The reactivity worth of top beryllium shim data plates was calculated to be 19.072 mk. The result of depletion analysis for NIRR-1 core shows that (7.9947 ± 0.0008) g of U-235 was consumed for the period of 12 years of operating time. The production of the build-up of Pu-239 was found to be (0.0347 ± 0.0043) g. The core life time estimated in this research was found to be 30.33 years. This is in good agreement with the literature

  20. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

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    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  1. Burnup calculations using the OREST computer code for uranium dioxide fuel elements of boiling water reactors. Abbrandberechnung mit OREST fuer Urandioxid-Siedewasserreaktor-Brennelemente

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U.

    1991-01-01

    There are plans to also use plutonium containing fuel elements (mixed oxide fuel) in the BWR type reactors, with a proportion of up to one third of the entire fuel core. The new concept uses complete MOX fuel elements, as are used in the PWR type reactors. The OREST computer code has been designed for burnup calculations in PWRs. The situation in BWRs is different, as in these reactor types, fuel elements are heterogenous in design, and burnup calculations have to take into account the axial variations of the void fraction, so that multi-dimensional effects have to be calculated. The report explains that the one-dimensional OREST code can be enhanced by supplementing calculations, performed with the Monte-Carlo type KENO code in this case, and is thus suitable without restrictions for performing burnup calculations for MOX fuel elements in BWRs. The calculation method and performance is illustrated by the example of a UO{sub 2} fuel element of the Wuergassen reactor. The model calculations predict a relatively high residual activity in the upper part of the fuel element, and a distinct curium buildup in the lower third of the BWR fuel element. (orig./HP).

  2. Uncertainty Propagation Analysis for PWR Burnup Pin-Cell Benchmark by Monte Carlo Code McCARD

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    Ho Jin Park

    2012-01-01

    Full Text Available In the Monte Carlo (MC burnup analyses, the uncertainty of a tally estimate at a burnup step may be induced from four sources: the statistical uncertainty caused by a finite number of simulations, the nuclear covariance data, uncertainties of number densities, and cross-correlations between the nuclear data and the number densities. In this paper, the uncertainties of kinf, reaction rates, and number densities for a PWR pin-cell benchmark problem are quantified by an uncertainty propagation formulation in the MC burnup calculations. The required sensitivities of tallied parameters to the microscopic cross-sections and the number densities are estimated by the MC differential operator sampling method accompanied by the fission source perturbation. The uncertainty propagation analyses are conducted with two nuclear covariance data—ENDF/B-VII.1 and SCALE6.1/COVA libraries—and the numerical results are compared with each other.

  3. Simulation of Low-Enriched Uranium (LEU) Burnup in Russian VVER Reactors with the HELIOS Code Package

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.

    2000-03-01

    The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.

  4. Analyses of PWR spent fuel composition using SCALE and SWAT code systems to find correction factors for criticality safety applications adopting burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor

  5. Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

  6. Des macros Excel pour exporter et importer des modules de code VBA

    OpenAIRE

    Duclos, Jean Baptiste

    2008-01-01

    Cet article présente un outil Excel qui permet de gérer facilement l’exportation et l’importation de macros VBA, en sauvegardant les programmes des macros dans des fichiers textes. Les macros d’importation et d’exportation de code peuvent être facilement utilisées dans d’autres programmes.

  7. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  8. Mathematic preprocessor for RELAP5 code using Microsoft Excel; Pre-processador matematico para o codigo RELAP5 utilizando o Microsoft Excel

    Energy Technology Data Exchange (ETDEWEB)

    Paladino, Patricia Andrea

    2006-07-01

    Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a toll for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-Excel software. (author)

  9. Influence of FIMA burnup on actinides concentrations in PWR reactors

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2016-01-01

    Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.

  10. Fuel burnup measurements in FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Rawlins, J.A.; Wootan, D.W.; Dobbin, K.D.

    1984-08-01

    Fuel burnup and isotopic fission rates were measured in FFTF during acceptance testing in an 8.6 day full power irradiation. Results were compared with three-dimensional diffusion theory calculations based on ENDF/B-V cross sections. A bias of about 3% exists between burnup and fission rate data, although measured axial and radial profiles are in good agreement. The calculated and measured radial power distributions are in disagreement by 6 to 10% from core center to the outer row of fuel. At core center, calculation/experiment (C/E) for isotopic fission rates is generally 1.01 while C/E for burnup is 1.04. Overall measurement uncertainties are 3% and 2% for fission rate and burnup experiments, respectively. Application of the results to a long-range goal of calculating burnup to 1% accuracy is discussed.

  11. Flexible modified candle burnup scheme based long life Pb-Bi cooled fast reactor with natural uranium as fuel cycle input employing coupled core

    Energy Technology Data Exchange (ETDEWEB)

    Su' ud, Zaki; SNM, Rida [Physics Dept., ITB, Jl. Ganesha 10, Bandung, West Java 40132 (Indonesia); Sekimoto, Hiroshi [Tokyo Inst. of Technology (Japan)

    2009-06-15

    calculation is then brought back to SRAC code for cell burn-up calculation. This iteration is repeated until the convergence is reached. Using the coupling dual core concept after some optimization we got several design which can fulfill the criteria can be obtained. Among of the main results are the design of long-life Pb-Bi cooled fast reactors which fuel cycle input is natural uranium but the maximum burnup level can be adjusted from about 10% HM to 40% HM. The results show that the twin(coupling )core concept can work well and give excellent configuration of modified CANDLE with flexible maximum burnup level up which can reach current fast reactors fuel technology. References: 1. H. Sekimoto, Light a CANDLE An Innovative Burnup Strategy of Nuclear Reactors, COE INES, Research Lab. For Nuclear Reactors, Tokyo Institute of Technology, Japan. 2. H. Sekimoto, et al, 'CANDLE: The New Burnup Strategy', Nuclear Science and Engineering, 139, 1-12 (2001), 3. Zaki S and H. Sekimoto: 'Optimization of Modified Candle Burnup Scheme Based Long Life Pb-Bi Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input', PBNC 2008 Conference, Oct. 11-18, 2008, Aomori Japan. 4. Zaki S and H. Sekimoto: 'Design Study Of Long Life Pb-Bi Cooled Fast Reactor With Natural Uranium As Fuel Cycle Input Using Modified Candle Burnup Scheme' Submitted to Annals of Nuclear Energy. (authors)

  12. Designing Critical Experiments in Support of Full Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Roberts, Jeremy A [ORNL

    2008-01-01

    Burnup credit is the process of accounting for the negative reactivity due to fuel burnup and generation of parasitic absorbers over fuel assembly lifetime. For years, the fresh fuel assumption was used as a simple bound in criticality work for used fuel storage and transportation. More recently, major actinides have been included [1]. However, even this yields a highly conservative estimate in criticality calculations. Because of the numerous economical benefits including all available negative reactivity (i.e., full burnup credit) could provide [2], it is advantageous to work toward full burnup credit. Unfortunately, comparatively little work has been done to include non-major actinides and other fission products (FP) in burnup credit analyses due in part to insufficient experimental data for validation of codes and nuclear data. The Burnup Credit Criticality Experiment (BUCCX) at Sandia National Laboratory was a set of experiments with {sup 103}Rh that have relevance for burnup credit [3]. This work uses TSUNAMI-3D to investigate and adjust a BUCCX model to match isotope-specific, energy-dependent k{sub eff} sensitivity profiles to those of a representative high-capacity cask model (GBC-32) [4] for each FP of interest. The isotopes considered are {sup 149}Sm, {sup 143}Nd, {sup 103}Rh, {sup 133}Cs, {sup 155}Gd, {sup 152}Sm, {sup 99}Tc, {sup 145}Nd, {sup 153}Eu, {sup 147}Sm, {sup 109}Ag, {sup 95}Mo, {sup 150}Sm, {sup 101}Ru, and {sup 151}Eu. The goal is to understand the biases and bias uncertainties inherent in nuclear data, and ultimately, to apply these in support of full burnup credit.

  13. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    Directory of Open Access Journals (Sweden)

    Jung Suk Kim

    2015-12-01

    Full Text Available The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233U, 242Pu, 150Nd, and 133Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

  14. Analysis of the burnup of the control rods with the COREMASTER-Presto code; Analisis del quemado de barras de control con el codigo COREMASTER-PRESTO

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jlhm@nuclear.inin-mx

    2003-07-01

    An evaluation of the capacity of the COREMASTER-Presto code, to evaluate generically the burnt of the control bars in the Laguna Verde reactors plant (CLV) is made. It was found that the code only reports burnt values of the control rods in MWD/TM, in spite of having with a second order polynomial model, for the conversion to remainder of the Boron-10 (B-10). It was observed that said model is adequate only for burnt smaller to 45,000 MWD/TM. To evaluate the burnt of the control rods it was reproduced the balance cycle of 18 months for the CLV, executing Cm-Presto during 13 consecutive cycles. First without rod burnt, taking this as the base case. Later on, cases with 1, 2 and up to 13 cycles with rod burnt were generated. When comparing results it was observed that the control rods pattern it loses reactivity lineally with the burnt one. By each 10 G Wd/T of burnt of the nucleus it is decreased the reactivity of the pattern rods {approx} 1 pcm in hot condition and of {approx} 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)

  15. Dependence of control rod worth on fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Savva, P., E-mail: savvapan@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Varvayanni, M., E-mail: melina@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece); Catsaros, N., E-mail: nicos@ipta.demokritos.g [NCSR ' DEMOKRITOS' , PoB 60228, 15310 Aghia Paraskevi (Greece)

    2011-02-15

    Research highlights: Diffusion and MC calculations for rod worth dependence on burnup and Xe in reactors. One-step rod withdrawal/insertion are used for rod worth estimation. The study showed that when Xe is present the rods worth is significantly reduced. Rod worth variation with burnup depends on rod position in core. Rod worth obtained with MC code is higher than that obtained from deterministic. - Abstract: One important parameter in the design and the analysis of a nuclear reactor core is the reactivity worth of the control rods, i.e. their efficiency to absorb excess reactivity. The control rod worth is affected by parameters such as the fuel burnup in the rod vicinity, the Xe concentration in the core, the operational time of the rod and its position in the core. In the present work, two different computational approaches, a deterministic and a stochastic one, were used for the determination of the rods worth dependence on the fuel burnup level and the Xe concentration level in a conceptual, symmetric reactor core, based on the MTR fuel assemblies used in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system composed by the SCALE modules NITAWL and XSDRN and the diffusion code CITATION was used, while for the stochastic one the Monte Carlo code TRIPOLI was applied. The study showed that when Xe is present in the core, the rods worth is significantly reduced, while the rod worth variation with increasing burnup depends on the rods position in the core grid. The rod worth obtained with the use of the Monte Carlo code is higher than the one obtained from the deterministic code.

  16. Burnup Credit of Erbia Super-High-Burnup Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sugimura, Naoki; Imamura, Michitaka; Mori, Masaaki [Nuclear Engineering, Ltd., Osaka (Japan); Yamasaki, Masatoshi [Nuclear Fuel Industries, Ltd., Osaka (Japan)

    2008-07-01

    Based on the concept of the Erbia bearing Super High-Burnup (Er-SHB) fuel, the initial erbia contents to guarantee the lower reactivity than that of the conventional 5.0 wt% enriched UO{sub 2} fuels during burnup and cooling are studied. According to the results, the feasibility of the commercial PWR cores using Er-SHB fuels is verified. As the results, it is verified that the long life core operation using Er-SHB fuel are feasible and approximately 20% of feed fuel assemblies can be saved by using Er-SHB fuel. (authors)

  17. Performance limitations for PWR fuel burnup extensions and high burnup complex loading patterns; Esquemas de recarga de combustible de alto quemado y limitaciones asociadas

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Aragones, J. M.; Cabellos, O.; Garcia Herranz, N. [Universidad Politecnica de Madrid (Spain)

    2000-07-01

    The analysis, design and on-line surveillance of pressurized water reactors require extensive and detailed 3D core calculations. The development and the improvement of codes are required due to the increasing heterogeneity in PWR (new type of fuel assemblies, complex loading patterns and safety and reliability requirements of nuclear reactor operation). The SEANAP system has been extensively validated and this system has been used to analyze high burnup fuel specifications and several types of long cycles. Some Programs (with Spanish participation) have been developed to study limitations for fuel burnup extensions: Robust Fuel Program, Segmented Rods Program and gain regulatory acceptance of fuel design and operation to higher burnup levels. (Author)

  18. Criticality Evaluation of GBC-32 Cask with HBN No.3 Fuels in PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do-Yeon; Yoon, Hyoungju; Park, Kwangheon; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

    2015-10-15

    An application of burnup credit is able to increase the capacity in casks. In this paper, the criticality evaluation for burnup credit was performed for the GBC-32 cask with the fuel assemblies discharged after HBN No.3 Cycle 6 by SCALE6.1/STARBUCS and MCNP6 with the axial burnup distributions and average discharge burnups evaluated using DeCART and MASTER codes. The criticality evaluation for burnup credit was performed for the GBC-32 cask with the fuel assemblies discharged after HBN No.3 Cycle 6 by STARBUCS and MCNP6 codes with the axial burnup distributions and average discharge burnups evaluated using DeCART and MASTER codes. k{sub eff} values and end effects were calculated for 3 cooling times of 0, 20, and 30 years. From the results calculated in these conditions, the following conclusions are drawn. (1) 12 discharged fuel assemblies for the cooling time of 0 year were not allowed to be stored in the cask because the estimated k{sub eff} values exceeds 0.9146. (2) Most of the discharged fuel assemblies except for 3 discharged fuel assemblies were allowed to be stored for the cooling times of 20 and 30 years. (3) The end effects increased as the cooling time increases, within the maximums of 834.93 pcm for the cooling time of 0 year, 1684.45 pcm for 20 years, and 2178.92 pcm for 30 years.

  19. Validation of lattice code 'EXCEL' with TIC experiments on uniform and regularly perturbed lattices

    Energy Technology Data Exchange (ETDEWEB)

    Ramakrishna, A., E-mail: anantatmula.ramakrishna@gmail.co [Atomic Energy Regulatory Board, Niyamak Bhavan, Anushaktinagar, Mumbai 400 094 (India); Jagannathan, V. [Light Water Reactors Physics Section, Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Jain, R.P. [IIT Bombay, Mumbai (India)

    2010-12-15

    Temporary International Collective (TIC) was established in 1972 by an agreement among seven countries, namely, Bulgaria, Czechoslovakia, Germany, Hungary, Poland, Romania and Union of Soviet Socialist Republics. The main objective of TIC was to provide the experimental data for the reactor physics analysis of water cooled and water moderated power reactors (WWER). Extensive experimental work for different core configurations was carried out by TIC countries to investigate the physics behaviour of WWER lattices and the results were published in TIC volumes. Two VVER-1000 MWe reactors are currently in an advanced stage of construction and due for commissioning in Kudankulam, Tamil Nadu, India. Indigenous development of in-core fuel management computer codes for the analysis of hexagonal lattice cores is also in an advanced stage to address various design, operation and safety issues of VVER type cores. The validation of the above TIC lattice experiments will help in the identification of deficiencies in reactor physics design computational codes and the associated nuclear data libraries. In this paper, TIC experiments on uniform and regularly perturbed lattices have been analyzed as part of the validation of indigenous computer codes, EXCEL, TRIHEX-FA and HEXPIN developed at Light Water Reactors Physics Section, B.A.R.C. Neutron-nuclear multi-group cross-section libraries in WIMS/D format in 69/172 energy groups have been released by IAEA at the conclusion of WIMS library update project (WLUP). In the present study we have used libraries based on ENDF/B-6, ENDF/B-7, JEFF3.1 and JENDL3.2 evaluated nuclear datasets. The results of the theoretical analyses bring out the performance of the code system and various cross-section libraries.

  20. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  1. 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.

  2. Using Microsoft Excel as a pre-processor for CODE V optimization of air spaces when building camera lenses

    Science.gov (United States)

    Stephenson, Dave

    2013-09-01

    When building high-performance camera lenses, it is often preferable to tailor element-to-element air spaces instead of tightening the fabrication tolerances sufficiently so that random assembly is possible. A tailored air space solution is usually unique for each serial number camera lens and results in nearly nominal performance. When these air spaces are computed based on measured radii, thickness, and refractive indices, this can put a strain on the design engineering department to deal with all the data in a timely fashion. Excel† may be used by the assembly technician as a preprocessor tool to facilitate data entry and organization, and to perform the optimization using CODE V‡ (or equivalent) without any training or experience in using lens design software. This makes it unnecessary to involve design engineering for each lens serial number, sometimes waiting in their work queue. In addition, Excel can be programmed to run CODE V in such a way that discrete shim thicknesses result. This makes it possible for each tailored air space solution to be achieved using a finite number of shims that differ in thickness by a reasonable amount. It is generally not necessary to tailor the air spaces in each lens to the micron level to achieve nearly nominal performance.

  3. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  4. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  5. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  6. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  7. A criticality analysis of the GBC-32 dry storage cask with Hanbit nuclear power plant unit 3 fuel assemblies from the viewpoint of burnup credit

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Hyung Ju; Kim, Do Yeon; Park, Kwang Heon; Hong, Ser Gi [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

    2016-06-15

    Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed for the GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit Nuclear Power Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributions and average discharge burnups were evaluated using the DeCART and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs were performed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit using SCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). The axial burnup distributions were determined for 20 UNFAs with various initial enrichments and burnups, which were applied to the criticality analysis for the cask system. The UNFAs for 20- and 30-year cooling times were assumed to be stored in the cask. The criticality analyses indicated that keff values for UNFAs with nonuniform axial burnup distributions were larger than those with a uniform distribution, that is, the end effects were positive but much smaller than those with the reference distribution. The axial burnup distributions for 20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upper region than the reference ones of the United States Department of Energy. These differences in the axial burnup distributions resulted in a significant reduction in end effects compared with the reference.

  8. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Enercon Services, Inc.

    2011-03-14

    ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost

  9. CXTFIT/Excel-A modular adaptable code for parameter estimation, sensitivity analysis and uncertainty analysis for laboratory or field tracer experiments

    Science.gov (United States)

    Tang, Guoping; Mayes, Melanie A.; Parker, Jack C.; Jardine, Philip M.

    2010-09-01

    We implemented the widely used CXTFIT code in Excel to provide flexibility and added sensitivity and uncertainty analysis functions to improve transport parameter estimation and to facilitate model discrimination for multi-tracer experiments on structured soils. Analytical solutions for one-dimensional equilibrium and nonequilibrium convection dispersion equations were coded as VBA functions so that they could be used as ordinary math functions in Excel for forward predictions. Macros with user-friendly interfaces were developed for optimization, sensitivity analysis, uncertainty analysis, error propagation, response surface calculation, and Monte Carlo analysis. As a result, any parameter with transformations (e.g., dimensionless, log-transformed, species-dependent reactions, etc.) could be estimated with uncertainty and sensitivity quantification for multiple tracer data at multiple locations and times. Prior information and observation errors could be incorporated into the weighted nonlinear least squares method with a penalty function. Users are able to change selected parameter values and view the results via embedded graphics, resulting in a flexible tool applicable to modeling transport processes and to teaching students about parameter estimation. The code was verified by comparing to a number of benchmarks with CXTFIT 2.0. It was applied to improve parameter estimation for four typical tracer experiment data sets in the literature using multi-model evaluation and comparison. Additional examples were included to illustrate the flexibilities and advantages of CXTFIT/Excel. The VBA macros were designed for general purpose and could be used for any parameter estimation/model calibration when the forward solution is implemented in Excel. A step-by-step tutorial, example Excel files and the code are provided as supplemental material.

  10. Summary of high burnup fuel issues and NRC`s plan of action

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  11. A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Makmal, T. [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel); Nuclear Physics and Engineering Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Aviv, O. [Radiation Safety Division, Soreq Nuclear Research Center, Yavne 81800 (Israel); Gilad, E., E-mail: gilade@bgu.ac.il [The Unit of Nuclear Engineering, Ben-Gurion University of The Negev, Beer-Sheva 84105 (Israel)

    2016-10-21

    A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections. - Highlights: • Simple, inexpensive, safe and flexible experimental setup that can be quickly deployed. • Experimental results are thoroughly corroborated against ORIGEN2 burnup code. • Experimental uncertainty of 9% and 5% deviation between measurements and simulations. • Very high burnup MTR fuel element is examined, with 60% depletion of {sup 235}U. • Impact of highly irregular irradiation regime on burnup evaluation is studied.

  12. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile

    Energy Technology Data Exchange (ETDEWEB)

    Barkauskas, V., E-mail: vytenis.barkauskas@ftmc.lt; Plukiene, R., E-mail: rita.plukiene@ftmc.lt; Plukis, A., E-mail: arturas.plukis@ftmc.lt

    2016-10-15

    Highlights: • RBMK-1500 fuel burn-up impact on k{sub eff} in the SNF cask was calculated using SCALE 6.1. • Positive end effect was noticed at certain burn-up for the RBMK-1500 spent nuclear fuel. • The non-uniform uranium depletion is responsible for the end effect in RBMK-1500 SNF. • k{sub eff} in the SNF cask does not exceed a value of 0.95 which is set in the safety requirements. - Abstract: Safe long-term storage of spent nuclear fuel (SNF) is one of the main issues in the field of nuclear safety. Burn-up credit application in criticality analysis of SNF reduces conservatism of usually used fresh fuel assumption and implies a positive economic impact for the SNF storage. Criticality calculations of spent nuclear fuel in the CONSTOR® RBMK-1500/M2 cask were performed using pre-generated ORIGEN-ARP spent nuclear fuel composition libraries, and the results of the RBMK-1500 burn-up credit impact on the effective neutron multiplication factor (k{sub eff}) have been obtained and are presented in the paper. SCALE 6.1 code package with the STARBUCKS burn-up credit evaluation tool was used for modeling. Pre-generated ARP (Automatic Rapid Processing) crosssection libraries based on ENDF/B-VII cross section library were used for fast burn-up inventory modeling. Different conditions in the SNF cask were modeled: 2.0% and 2.8% initial enrichment fuel of various burn-up and water density inside cavities of the SNF cask. The fuel composition for the criticality analysis was chosen taking into account main actinides and most important fission products used in burn-up calculations. A significant positive end effect is noticed from 15 GWd/tU burn-up for 2.8% enrichment fuel and from 9 GWd/tU for 2.0% enrichment fuel applying the actinide-only approach. The obtained results may be applied in further evaluations of the RBMK type reactor SNF storage as well as help to optimize the SNF storage volume inside the CONSTOR® RBMK-1500/M2 cask without compromising criticality

  13. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  14. Technical Basis for Peak Reactivity Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (keff) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup

  15. NUCLEAR DATA UNCERTAINTY PROPAGATION FOR A TYPICAL PWR FUEL ASSEMBLY WITH BURNUP

    Directory of Open Access Journals (Sweden)

    D. ROCHMAN

    2014-06-01

    Full Text Available The effects of nuclear data uncertainties are studied on a typical PWR fuel assembly model in the framework of the OECD Nuclear Energy Agency UAM (Uncertainty Analysis in Modeling expert working group. The “Fast Total Monte Carlo” method is applied on a model for the Monte Carlo transport and burnup code SERPENT. Uncertainties on k∞, reaction rates, two-group cross sections, inventory and local pin power density during burnup are obtained, due to transport cross sections for the actinides and fission products, fission yields and thermal scattering data.

  16. Spent fuel pool storage calculations using the ISOCRIT burnup credit tool

    Energy Technology Data Exchange (ETDEWEB)

    Kucukboyaci, Vefa [Westinghouse Electric Company, Cranberry Township, PA; Marshall, William BJ J [ORNL

    2012-01-01

    In order to conservatively apply burnup credit in spent fuel pool criticality safety analyses, Westinghouse has developed a software tool, ISOCRIT, for generating depletion isotopics. This tool is used to create isotopics data based on specific reactor input parameters, such as design basis assembly type; bounding power/burnup profiles; reactor specific moderator temperature profiles; pellet percent theoretical density; burnable absorbers, axial blanket regions, and bounding ppm boron concentration. ISOCRIT generates burnup dependent isotopics using PARAGON; Westinghouse's state-of-the-art and licensed lattice physics code. Generation of isotopics and passing the data to the subsequent 3D KENO calculations are performed in an automated fashion, thus reducing the chance for human error. Furthermore, ISOCRIT provides the means for responding to any customer request regarding re-analysis due to changed parameters (e.g., power uprate, exit temperature changes, etc.) with a quick turnaround.

  17. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  18. Microhardness and Young's modulus of high burn-up UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cappia, F. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125, Karlsruhe (Germany); Technische Universität München, Faculty of Mechanical Engineering, Department of Nuclear Engineering, D-85748, Garching bei München (Germany); Pizzocri, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125, Karlsruhe (Germany); Politecnico di Milano, Department of Energy, Nuclear Engineering Division, 20156, Milano (Italy); Marchetti, M. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125, Karlsruhe (Germany); Université Montpellier 2, Institut d’Electronique du Sud UMR CNRS 5214, 34095, Montpellier (France); Schubert, A.; Van Uffelen, P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125, Karlsruhe (Germany); Luzzi, L. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, 20156, Milano (Italy); Papaioannou, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125, Karlsruhe (Germany); Macián-Juan, R. [Technische Universität München, Faculty of Mechanical Engineering, Department of Nuclear Engineering, D-85748, Garching bei München (Germany); Rondinella, V.V., E-mail: Vincenzo.RONDINELLA@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125, Karlsruhe (Germany)

    2016-10-15

    Vickers microhardness (HV{sub 0.1}) and Young's modulus (E) measurements of LWR UO{sub 2} fuel at burn-up ≥60 GWd/tHM are presented. Their ratio HV{sub 0.1}/E was found constant in the range 60–110 GWd/tHM. From the ratio and the microhardness values vs porosity, the Young's modulus dependence on porosity was derived and extended to the full radial profile, including the high burn-up structure (HBS). The dependence is well represented by a linear correlation. The data were compared to fuel performance codes correlations. A burn-up dependent factor was introduced in the Young's modulus expression. The modifications extend the experimental validation range of the TRANSURANUS correlation from un-irradiated to irradiated UO{sub 2} and up to 20% porosity. First simulations of LWR fuel rod irradiations were performed in order to illustrate the impact on fuel performance. In the specific cases selected, the simulations suggest a limited effect of the Young's modulus decrease due to burn-up on integral fuel performance. - Highlights: • Vickers microhardness and Young's modulus data of high burnup fuels are presented. • The data are compared to fuel performance codes' correlations. • A burn-up dependent factor is introduced for the Young's modulus of irradiated fuel. • The modification extends ranges of experimental validation of the code correlation. • The new burn-up dependent factor has limited effect on integral fuel performance.

  19. Impact investigation of reactor fuel operating parameters on reactivity for use in burnup credit applications

    Science.gov (United States)

    Sloma, Tanya Noel

    When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light

  20. AREVA NP burnup credit investigation on irradiated MOX fuel within the REBUS BWR programme

    Energy Technology Data Exchange (ETDEWEB)

    Alander, Alexandra; Misu, Stefan; Timm, Wolf [AREVA, AREVA NP, Erlangen (Germany); Thareau, Sebastien [AREVA, AREVA NP, Paris (France)

    2008-07-01

    The present paper summarizes a criticality and burnup credit investigation carried out using the 2D spectral codes CASMO-4 and APOLLO2-A. Fission rate distributions and multiplication factors, for UOX and MOX configurations, are calculated as well as the reactivity effect caused by burnup on a selection of irradiated MOX fuel assemblies. 3D core criticality calculations were carried out with the Monte Carlo transport code MOCA and the deterministic transport code VARIANT (a nodal code developed by ANL) using CASMO-4 generated cross section libraries. Calculations were compared to experimental data from the critical facility VENUS in the context of the REBUS BWR Programme. The results confirm that the spectral codes CASMO-4 and APOLLO2-A are well suited to calculate fission rates, multiplication factors and reactivity effects. It is also found that the calculated burnup reactivity effect, using CASMO-4 generated cross sections and the best 3D method (MOCA), is underestimated by merely 7% compared to the experimental value, which can mainly be attributed to the simplifications done in order to model the critical configurations with reasonable efforts. (authors)

  1. Incorporation of the variation in conductivity with burnup in the stability of code predictive LAPUR; Incroporacion de la variacion de la conductividad con el quemado en el codigo de estabilidad predictivo LAPUR

    Energy Technology Data Exchange (ETDEWEB)

    Escriba, A.; Munoz-cobo, J. L.; Merino, R.; Melara, J.; Albendea, M.

    2013-07-01

    In the field of nuclear safety, the analysis of the stability of boiling water reactors is one of the biggest challenges for researchers. LAPUR code that allows to obtain the parameters of stability of the plant (Decay rate and frequency), being one of the programs used by IBERDROLA can be used for these calculations. With the collaboration of the research group TIN of the Polytechnic University of Valencia, a model of loss of conductivity of uranium has joined with the burned LAPUR. This update allows you to play the phenomenon in a more realistic way. This improvement has been validated and verified contrasting results with reference values.

  2. Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)

    2008-04-15

    Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.

  3. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

  4. Simulation of triton burn-up in JET plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Loughlin, M.J.; Balet, B.; Jarvis, O.N.; Stubberfield, P.M. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-07-01

    This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.

  5. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  6. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  7. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    Energy Technology Data Exchange (ETDEWEB)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Suga, Masataka [Kokan Keisoku Co., Kawasaki, Kanagawa (Japan)

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  8. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL

    2016-01-01

    A technical basis for peak reactivity boiling water reactor (BWR) burnup credit (BUC) methods was recently generated, and the technical basis for extended BWR BUC is now being developed. In this paper, a number of effects related to extended BWR BUC are analyzed, including three major operational effects in BWRs: the coolant density axial distribution, the use of control blades during operation, and the axial burnup profile. Specifically, uniform axial moderator density profiles are analyzed and compared to previous results and an additional temporal fidelity study combing moderator density profiles for three different fuel assemblies is presented. Realistic control blade histories and cask criticality results are compared to previously generated constructed control blade histories. Finally, a preliminary study of the axial burnup profile is provided.

  9. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  10. Benefits of the delta K of depletion benchmarks for burnup credit validation

    Energy Technology Data Exchange (ETDEWEB)

    Lancaster, D. [NuclearConsultants.com, 187 Faith Circle, Boalsburg, PA 16827 (United States); Machiels, A. [Electric Power Research Inst., Inc., 3420 Hillview Avenue, Palo Alto, CA 94304 (United States)

    2012-07-01

    Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO{sub 2} critical experiments to determine the criticality safety limits on the neutron multiplication factor, k{sub eff}. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

  11. Study on nuclear physics of high burn-up full MOX-BWR core

    Energy Technology Data Exchange (ETDEWEB)

    Shirakawa, Toshihisa; Okubo, Tsutomu; Ochiai, Masa-aki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-08-01

    In this report, neutronics study of full Mixed-oxide (MOX) high burn-up BWR core is presented. Our design goals are about 3-year cycle length, four-batch refueling scheme and more than 100GWd/t fuel discharge burn-up. Base core configuration is 1,350MWe US version of ABWR with 9 x 9 type fuel assembly. Investigation of the reactor core has been carried out by arranging Gd{sub 2}O{sub 3} contents in fuel rods and changing water to fuel volume ratio (V{sub m}/V{sub f}) through the number of water rods or adjustment of fuel clad diameter. JAERI`s general purpose neutronics code system SRAC95 was used for two dimensional XY fuel assembly cell neutronics calculations. Calculated cases are for a comparatively high moderated fuel assembly with 9 water rods, a fuel assembly without water rods and a comparatively low moderated fuel assembly without water rods and with larger fuel clad diameter. All these 3 cases seem to achieve our design goals mentioned above. For the last case, three dimensional core burn-up calculation was performed by this code system. This case seems to attain a low linear power density and the operation with all control rod out. (author)

  12. The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.

  13. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  14. Effect of Control Blade History, and Axial Coolant Density and Burnup Profiles on BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-05-01

    Oak Ridge National Laboratory (ORNL) and the US Nuclear Regulatory Commission (NRC) have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation systems (often referred to as casks) and spent fuel pools (SFPs). This work is divided into two main phases. The first phase investigated the applicability of peak reactivity methods currently used in SFPs to transportation and storage casks and the validation of reactivity calculations and spent fuel compositions within these methods. The second phase focuses on extending BUC beyond peak reactivity. This paper documents the analysis of the effects of control blade insertion history, and moderator density and burnup axial profiles for extended BWR BUC.

  15. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    BSC

    2004-12-01

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  16. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  17. Underestimation of nuclear fuel burnup – theory, demonstration and solution in numerical models

    Directory of Open Access Journals (Sweden)

    Gajda Paweł

    2016-01-01

    Full Text Available Monte Carlo methodology provides reference statistical solution of neutron transport criticality problems of nuclear systems. Estimated reaction rates can be applied as an input to Bateman equations that govern isotopic evolution of reactor materials. Because statistical solution of Boltzmann equation is computationally expensive, it is in practice applied to time steps of limited length. In this paper we show that simple staircase step model leads to underprediction of numerical fuel burnup (Fissions per Initial Metal Atom – FIMA. Theoretical considerations indicates that this error is inversely proportional to the length of the time step and origins from the variation of heating per source neutron. The bias can be diminished by application of predictor-corrector step model. A set of burnup simulations with various step length and coupling schemes has been performed. SERPENT code version 1.17 has been applied to the model of a typical fuel assembly from Pressurized Water Reactor. In reference case FIMA reaches 6.24% that is equivalent to about 60 GWD/tHM of industrial burnup. The discrepancies up to 1% have been observed depending on time step model and theoretical predictions are consistent with numerical results. Conclusions presented in this paper are important for research and development concerning nuclear fuel cycle also in the context of Gen4 systems.

  18. Excel simulations

    CERN Document Server

    Verschuuren, Gerard M

    2013-01-01

    Covering a variety of Excel simulations, from gambling to genetics, this introduction is for people interested in modeling future events, without the cost of an expensive textbook. The simulations covered offer a fun alternative to the usual Excel topics and include situations such as roulette, password cracking, sex determination, population growth, and traffic patterns, among many others.

  19. Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.

  20. Conservative approach for PWR MOX Burnup Credit implementation

    Energy Technology Data Exchange (ETDEWEB)

    Jutier, Ludyvine; Checiak, Benoit; Raby, Jerome; Aguiar, Luis; Le Bars, Igor [IRSN, Fontenay-aux-Roses (France)

    2008-07-01

    Burnup Credit allows considering the reactivity decrease due to fuel irradiation in criticality studies for the nuclear fuel cycle. Its implementation requires to carefully analyze the validity of the assumption made to: define the axial profile of the burnup, determine the composition of the irradiated fuel and compute the criticality simulation. In the framework of Burnup Credit implementation for PWR mixed oxide fuels (MOX), this paper focus on the determination of a conservative inventory of the irradiated fuel. The studies presented in this paper concern: the influence of irradiation conditions and of the MOX fuel initial composition on the irradiated MOX fuel reactivity. Criticality calculations are also performed for PWR MOX fuel industrial applications in order to get Burnup Credit gain estimations. (authors)

  1. A guide introducing burnup credit, preliminary version. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)

  2. Investigation of very high burnup UO{sub 2} fuels in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cappia, Fabiola

    2017-03-27

    fuel mechanical properties and their relationship with the local microstructure at high burnup has been recognised, being one of the factors influencing Pellet-Cladding Mechanical Interaction (PCMI). The knowledge of the fuel mechanical properties has also fundamental importance to assess the mechanical integrity of the spent fuel during the back end of the fuel cycle. In this context, the scope of this work was twofold. The first task was the experimental study of the fuel microhardness and Young's modulus in high burnup UO{sub 2} fuels and their relationship with the local porosity, which has a major impact on their variation. Moreover, assessment of the accumulation of the decay damage during storage and its influence on the fuel microhardness has been carried out, in the framework of safety studies on the back end of the fuel cycle at high burnup. The second task consisted in the evaluation of the porosity and pore size distribution evolution in high burnup fuel, with particular focus on the HBS porosity. The experimental relationship between the high burnup fuel Young's modulus and local porosity obtained through combination of acoustic microscopy and microindentation measurements has been compared to the material property correlations commonly used in fuel performance codes, which are based on data from characterization of unirradiated UO{sub 2}. The investigation has revealed that the relationship is similar for non-irradiated and irradiated material, but in the latter case an additional factor that takes into account the Young's modulus decrease due to burnup accumulation has to be included in the correlation to match the experimental values. First analysis of the fuel microhardness as a function of the accumulated decay damage has shown that fuel microhardness does not significantly increase when the dose due to the additional decay damage accumulated during storage reaches ∼ 0.1 dpa, in agreement with what observed in unirradiated {sup 238}Pu

  3. Extended burnup core management for once-through uranium fuel cycles in LWRS. First annual report for the period 1 July 1979-30 June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Sesonske, A.

    1980-08-01

    Detailed core management arrangements are developed requiring four operating cycles for the transition from present three-batch loading to an extended burnup four-batch plan for Zion-1. The ARMP code EPRI-NODE-P was used for core modeling. Although this work is preliminary, uranium and economic savings during the transition cycles appear of the order of 6 percent.

  4. Why Excel?

    Science.gov (United States)

    Barreto, Humberto

    2015-01-01

    This article is not the usual Excel pedagogy fare in that it does not provide an application or example taught via a spreadsheet. Instead, it briefly reviews the history of spreadsheets in the economics classroom and explores the current environment, with an emphasis on modern learning theory. The conclusion is not surprising: spreadsheets improve…

  5. SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Wagner, John C [ORNL

    2009-01-01

    The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality safety calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate system similarity on an integral and nuclide-reaction specific level. The results indicate that, except for the fresh fuel core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to a generic cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. Based on the integral system parameter, C{sub k}, approximately 30 of the 40 CRC state-points are applicable to validation of burnup credit in the generic cask containing typical spent fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. The state-points providing the highest similarity (C{sub k} > 0.95) were attained at or near the end of a reactor cycle. The C{sub k} values are dominated by neutron reactions with major actinides and hydrogen, as the sensitivities of these reactions are much higher than those of the minor actinides and fission products. On a nuclide-reaction specific level, the CRC state-points provide significant similarity for most of the actinides and fission products relevant to burnup credit. A comparison of energy-dependent sensitivity profiles shows a slight shift of the CRC K{sub eff} sensitivity profiles toward higher energies in the thermal region as compared to the K{sub eff} sensitivity profile of the generic cask

  6. Sourcing Excellence

    DEFF Research Database (Denmark)

    Adeyemi, Oluseyi

    2011-01-01

    the competitiveness of manufacturing companies (7). Over the last couple of years, supplier’s development is an evolution in supply chain management and there is a growing interest in generating approaches for meaningful development of suppliers and SDI. This would allow businesses to snatch long-term strategic......Sourcing Excellence is one of the key performance indicators (KPIs) in this world of ever changing sourcing strategies. Manufacturing companies need to access and diagnose the reliability and competencies of existing suppliers in order to coordinate and develop them. This would help in managing....... In essence, this would allow the deployment of both internal and external skills and competencies to determine the new approaches towards organizing, supporting and improving supplier driven innovations (SDI) and supplier relationship management (SRM). This is to achieve operational excellence...

  7. The interest of burnup increase in a context of recycling

    Energy Technology Data Exchange (ETDEWEB)

    Druenne, Hubert [GDF SUEZ-TRACTEBEL, Avenue Ariane 7 - B-1200 Brussels (Belgium)

    2009-06-15

    The current trend is to increase the fuel discharge burnup. In the framework of a recycling policy (closed cycle) higher burnup also affects the quality of the fissile material coming back from the reprocessing. It has the following consequences: - the lower quality of the reprocessed material requires either higher ERU enrichment and/or higher plutonium content in MOX fuel; - should some limits be reached (manufacture limits, maximum {sup 235}U enrichment or Pu content), the energy equivalence could no longer be maintained between recycled fuel assemblies (ERU or MOX) and ENU ones; - in turn, the loss of energy equivalence would request larger feed size, and hence would limit the burnup increase. Consequently, the question is whether an increase in burnup could hamper a recycling policy. In a closed cycle, and considering only the 2. or 3. generation PWR, does the increase in burnup still make it possible to reduce the need for fissile material? Is there an economic optimum? This paper attempts to answer these questions. In conclusions: A. Reprocessing and recycling processes: The present industrial processes technical limits do not hinder reprocessing and recycling of highly burned assemblies (at least in the assumed limit of roughly 65 GWd/t - assembly average). But taking into account the present limitations in terms of initial enrichment, enrichment technology, fuel fabrication and fresh fuel transportation, the additional costs may limit or even over-compensate the benefits resulting from higher discharge burnups. These costs must be known to conclude on the economy of high burnup in a closed cycle policy. B. In-core fuel management: In 18-month cycle, the present 5% enrichment limit is a strong constraint for 'auto-recycling': even with the shortest reprocessing delays as possible, it restricts the achievable batch average discharge burnup to about 56 GWd/t. The economy of high burnup should be analysed in a multi nuclear unit fuel management, with

  8. Addressing Fission Product Validation in MCNP Burnup Credit Criticality Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Bowen, Douglas G [ORNL; Marshall, William BJ J [ORNL

    2015-01-01

    The US Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation issued Interim Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC staff members’ review of burnup credit (BUC) analyses supporting transport and dry storage of pressurized water reactor spent nuclear fuel (SNF) in casks. The ISG includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MAs). Based on previous work documented in NRC Regulatory Guide (NUREG) Contractor Report (CR)-7109, the ISG recommends that NRC staff members accept the use of either 1.5 or 3% of the FP&MA worth—in addition to bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF—to conservatively account for the bias and bias uncertainty associated with the specified unvalidated FP&MAs. The ISG recommends (1) use of 1.5% of the FP&MA worth if a modern version of SCALE and its nuclear data are used and (2) 3% of the FP&MA worth for well qualified, industry standard code systems other than SCALE with the Evaluated Nuclear Data Files, Part B (ENDF/B),-V, ENDF/B-VI, or ENDF/B-VII cross sections libraries. The work presented in this paper provides a basis for extending the use of the 1.5% of the FP&MA worth bias to BUC criticality calculations performed using the Monte Carlo N-Particle (MCNP) code. The extended use of the 1.5% FP&MA worth bias is shown to be acceptable by comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII–based nuclear data. The comparison supports use of the 1.5% FP&MA worth bias when the MCNP code is used for criticality calculations, provided that the cask design is similar to the hypothetical generic BUC-32 cask model and that the credited FP&MA worth is no more than 0.1 Δkeff (ISG-8, Rev. 3, Recommendation 4).

  9. Advances in Metallic Fuels for High Burnup and Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.

    2016-10-01

    Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.

  10. Programming Excel services

    CERN Document Server

    Bruney, Alvin J

    2007-01-01

    Learn how to use Excel Services to add spreadsheets and workbooks to your enterprise dashboards and portals. You get real-world examples and code samples to help you enable business intelligence on Microsoft Office SharePoint Server 2007 applications-and simplify information sharing and reuse. Discover how to: Configure Windows SharePoint Products and Technologies, and provision Excel ServicesBuild and debug Web Parts derived from Microsoft .NET Framework or SharePoint base classes Customize and embed objects into Web PartsExtract workbook contents and bind the results to user interface con

  11. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Isotopic Composition Predictions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Wagner, John C [ORNL

    2011-01-01

    The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for

  12. Monte Carlo calculations of the REBUS critical experiment for validation of burn-up credit

    Energy Technology Data Exchange (ETDEWEB)

    Hennebach, M.; Kuhl, H. [WTI GmbH, Julich (Germany)

    2008-07-01

    The REBUS experiment is a valuable benchmark for the validation of Monte Carlo criticality codes within the context of burn-up credit. It investigates the difference in reactivity worth of unirradiated fuel rods and of fuel rods irradiated in a PWR. This paper presents results of criticality calculations for this experiment with the Monte Carlo codes MCNP and SCALE/KENO. The results show that even the comparatively small reactivity loss of about 2000 pcm measured in this experiment can be calculated within an accuracy of 6 per cent. A detailed comparison of measured and calculated core characteristics (fission rate and flux distributions) shows good agreement, which indicates an adequately detailed model geometry for criticality calculations. The slight underestimation (500 pcm) of absolute k(eff) values is not out of the typical range for criticality benchmark experiments.

  13. Impact of Reactor Operating Parameters on Cask Reactivity in BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Betzler, Benjamin R [ORNL; Ade, Brian J [ORNL

    2017-01-01

    This paper discusses the effect of reactor operating parameters used in fuel depletion calculations on spent fuel cask reactivity, with relevance for boiling-water reactor (BWR) burnup credit (BUC) applications. Assessments that used generic BWR fuel assembly and spent fuel cask configurations are presented. The considered operating parameters, which were independently varied in the depletion simulations for the assembly, included fuel temperature, bypass water density, specific power, and operating history. Different operating history scenarios were considered for the assembly depletion to determine the effect of relative power distribution during the irradiation cycles, as well as the downtime between cycles. Depletion, decay, and criticality simulations were performed using computer codes and associated nuclear data within the SCALE code system. Results quantifying the dependence of cask reactivity on the assembly depletion parameters are presented herein.

  14. Writing Excel Macros with VBA

    CERN Document Server

    Roman, Steven

    2008-01-01

    To achieve the maximum control and flexibility from Microsoft® Excel often requires careful custom programming using the VBA (Visual Basic for Applications) language. Writing Excel Macros with VBA, 2nd Edition offers a solid introduction to writing VBA macros and programs, and will show you how to get more power at the programming level: focusing on programming languages, the Visual Basic Editor, handling code, and the Excel object model.

  15. Passing excellence

    Science.gov (United States)

    Tsoupikova, Daria

    2007-02-01

    This paper describes the research and development of a virtual reality visualization project "Passing excellence" about the world famous architectural ensemble "Kizhi". The Kizhi Pogost is located on an island in Lake Onega in northern Karelia in Russia. It is an authentic museum of an ancient wood building tradition which presents a unique artistic achievement. This ensemble preserves a concentration of masterpieces of the Russian heritage and is included in the List of Most Endangered Sites of the World Monuments Watch protected by World Heritage List of UNESCO. The project strives to create a unique virtual observation of the dynamics of the architectural changes of the museum area beginning from the 15th Century up to the 21st Century. The visualization is being created to restore the original architecture of Kizhi island based on the detailed photographs, architectural and geometric measurements, textural data, video surveys and resources from the Kizhi State Open-Air Museum archives. The project is being developed using Electro, an application development environment for the tiled display high-resolution graphics visualization system and can be shown on the virtual reality systems such as the GeoWall TM and the C-Wall.

  16. Automated system for determining the burnup of spent nuclear fuel

    Directory of Open Access Journals (Sweden)

    Mokritskii V. A.

    2014-12-01

    Full Text Available The authors analyze their experience in application of semi-conductor detectors and development of a breadboard model of the monitoring system for spent nuclear fuel (SNF. Such system should use CdZnTe-detectors in which one-charging gathering conditions are realized. The proposed technique of real time SNF control during reloading technological operations is based on the obtained research results. Methods for determining the burnup of spent nuclear fuel based on measuring the characteristics of intrinsic radiation are covered in many papers, but those metods do not usually take into account that the nuclear fuel used during the operation has varying degrees of initial enrichment, or a new kind of fuel may be used. Besides, the known methods often do not fit well into the existing technology of fuel loading operations and are not suitable for operational control. Nuclear fuel monitoring (including burnup determination system in this research is based on the measurement of the spectrum of natural gamma-radiation of irradiated fuel assemblies (IFA, as from the point of view of minimizing the time spent, the measurement of IFA gamma spectra directly during fuel loading is optimal. It is the overload time that is regulated rather strictly, and burnup control operations should be coordinated with the schedule of the fuel loading. Therefore, the real time working capacity of the system should be chosen as the basic criterion when constructing the structure of such burnup control systems.

  17. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  18. 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)

  19. Impacts of burnup-dependent swelling of metallic fuel on the performance of a compact breed-and-burn fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Heo, Woong; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The U-Zr or U-TRU-Zr cylindrical metallic fuel slug used in fast reactors is known to swell significantly and to grow during irradiation. In neutronics simulations of metallic-fueled fast reactors, it is assumed that the slug has swollen and contacted cladding, and the bonding sodium has been removed from the fuel region. In this research, a realistic burnup-dependent fuel-swelling simulation was performed using Monte Carlo code McCARD for a single-batch compact sodium-cooled breed-and-burn reactor by considering the fuel-swelling behavior reported from the irradiation test results in EBR-II. The impacts of the realistic burnup-dependent fuel swelling are identified in terms of the reactor neutronics performance, such as core lifetime, conversion ratio, axial power distribution, and local burnup distributions. It was found that axial fuel growth significantly deteriorated the neutron economy of a breed-and-burn reactor and consequently impaired its neutronics performance. The bonding sodium also impaired neutron economy, because it stayed longer in the blanket region until the fuel slug reached 2% burnup.

  20. Modelling of fission gas swelling in the high burnup UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho

    1999-06-01

    Discharge burnup of the fuel in LWR has been increased to improve the fuel economy, and currently the high burnup fuel of over 70 MWd/kg U-rod avg. is being developed by the fuel vendors worldwide. At high burnup, thermal / mechanical properties of the fuel is known to change and new phenomenon could arise. This report describes the model development on fission gas swelling in high burnup UO{sub 2} fuel. For the low burnup fuel, swelling only by the solid fission products has been considered in the fuel performance analysis. However, at high burnup fuel, swelling by fission gas bubbles can not be neglected anymore. Therefore, fission gas swelling model which can predictbubble swelling of the high burnup UO{sub 2} fuel during the steady-state and the transient conditions in LWR was developed. Based on the bubble growth model, the empirical fission gas swelling model was developed as function of burnup, time and temperature. The model showed that fuel bubble swelling would be proportional to the burnup by the power of 1.157 and to the time by the power of 0.157. Comparison of the model prediction with the measured fission gas swelling data under the various burnup and temperature conditions showed that the model would predict the measured data reasonably well. (author). 20 refs., 8 tabs., 17 figs.

  1. Development of burnup dependent fuel rod model in COBRA-TF

    Science.gov (United States)

    Yilmaz, Mine Ozdemir

    The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN

  2. Burnup Credit Approach Used in the Yucca Mountain License Application

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [ORNL; Wagner, John C [ORNL

    2010-01-01

    The United States Department of Energy has submitted a license application (LA) for construction authorization of a deep geologic repository at Yucca Mountain, Nevada. The license application is currently under review by the United States Nuclear Regulatory Commission (NRC). This paper will describe the methodology and approach used in the LA to address the issue of criticality and the role of burnup credit during the postclosure period. The most significant and effective measures for prevention of criticality in the repository include multiple redundant barriers that act to isolate fissionable material from water (which can act as a moderator, corrosive agent, and transporter of fissile material); inherent geometry of waste package internals and waste forms; presence of fixed neutron absorbers in waste package internals; and fuel burnup for commercial spent nuclear fuel. A probabilistic approach has been used to screen criticality from the total system performance assessment. Within the probabilistic approach, criticality is considered an event, and the total probability of a criticality event occurring within 10,000 years of disposal is calculated and compared against the regulatory criterion. The total probability of criticality includes contributions associated with both internal (within waste packages) and external (external to waste packages) criticality for each of the initiating events that could lead to waste package breach. The occurrence of and conditions necessary for criticality in the repository have been thoroughly evaluated using a comprehensive range of parameter distributions. A simplified design-basis modeling approach has been used to evaluate the probability of criticality by using numerous significant and conservative assumptions. Burnup credit is used only for evaluations of in-package configurations and uses a combination of conservative and bounding modeling approximations to ensure conservatism. This paper will review the NRC regulatory

  3. Triton burnup study using scintillating fiber detector on JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1997-09-01

    The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

  4. Experiment Safety Assurance Package for the 40- to 52-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-hole Positions in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. T. Khericha; R. C. Pedersen

    2003-09-01

    This experiment safety assurance package (ESAP) is a revision of the last mixed uranium and plutonium oxide (MOX) ESAP issued in June 2002). The purpose of this revision is to provide a basis to continue irradiation up to 52 GWd/MT burnup [as predicted by MCNP (Monte Carlo N-Particle) transport code The last ESAP provided basis for irradiation, at a linear heat generation rate (LHGR) no greater than 9 kW/ft, of the highest burnup capsule assembly to 50 GWd/MT. This ESAP extends the basis for irradiation, at a LHGR no greater than 5 kW/ft, of the highest burnup capsule assembly from 50 to 52 GWd/MT.

  5. Excelling with Excel: Advanced Excel Functions for Collection Analysis

    OpenAIRE

    Pan, Denise; Wiersma, Gabrielle

    2014-01-01

    Microsoft Excel offers useful features and formulas that potentially allow acquisitions and collection development librarians to work smarter, not harder. Using journal cancellations as a workplace scenario, the presenters will provide attendees with ideas on how to organize data and complete basic calculations with Excel. The presenters provide examples on how to use several advanced Excel functions including PivotTables, VLOOKUP, and select formulas. They describe the steps for importing an...

  6. Development of a parallel processing couple for calculations of control rod worth in terms of burn-up in a WWER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering

    2017-03-15

    In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  7. Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  8. MOX fuel characterization for burnup credit application: Extension of nondestructive method qualified for LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Riffard, C.; Vidal, J.M. [Commissariat a l' Energie Atomique CEA/DEN/CAD, Building 230 Centre d tudes de Cadarache, 13108 St. Paul Lez Durance (France); Toubon, H. [COGEMA, 78 141 Velizy cedex (France); Pelletter, S. [Canberra-Eurisys, 78067 St. Quentin en Yvelines cedex (France); Batifol, M. [COGEMA La Hague, 50444 Beaumont Hague cedex (France)

    2006-07-01

    Before the reprocessing of low-enriched uranium (LEU) fuels at La. Hague plant, the assemblies are characterized with a nondestructive assay based on neutron emission (NE) and gamma-ray emission combined with the CESAR depletion code, giving the burnup (BU) with a good accuracy ({+-} 5% within a batch of fuels from one of COGEMA-La Hague's clients). The measurements confirm the hypothesis of the safety-criticality analysis of the process, in the context of the BU credit allowance. There is a need to extend the allowance of the reprocessing plants to the case of more highly enriched LEU fuels and to the case of mixed-oxide (MOX) fuels. The aim is to propose an upgraded method, valid for both LEU and MOX fuels, giving the average BU with an uncertainty lower than {+-} 15% for MOX fuels (without any modification of the current acceptance criteria for UO{sub 2} fuel, i.e. {+-} 15%), with a complementary module checking the operator data using the gamma-ray emission and the CESAR depletion code. In particular, the NE was interpreted with depletion calculations in the case of MOX fuels, which is the principal aim of this paper. This allows the BU determination of MOX fuels, which has been qualified during a measurement campaign in La Hague with 20 MOX assemblies. The mean BU of pressurized water reactor MOX assemblies has been determined for the first time with a maximum discrepancy of {+-} 5% compared to the declared value. (authors)

  9. Excel Initiative: Excellence in Youth Programming

    Directory of Open Access Journals (Sweden)

    Lynne M. Borden

    2015-06-01

    Full Text Available The Excellence in Youth Programming (Excel Initiative strives to support youth programs in delivering high quality programs. The backbone of Excel is the Youth Development Observational Tool (YDOT which allows for the virtual assessment of program staff who work with children and youth ages 9-18 years. The YDOT also allows Excel to provide structured feedback to programs. Excel has several unique features, including a virtual platform and a focus on the relationships between adults and youth participating in after-school programs. Offering structured assessment and interaction online eliminates expenses, provides convenient access for programs around the globe, and allows for unobtrusive assessment of worker-youth interactions. Excel is also integrated into a broader network of resources, tools, and research for those working with children and youth ages 9-18.

  10. Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, John C [ORNL; Parks, Cecil V [ORNL; Mueller, Don [ORNL; Gauld, Ian C [ORNL

    2010-01-01

    Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and

  11. Excel 2010 for dummies

    CERN Document Server

    Harvey, Greg

    2010-01-01

    The bestselling Excel book on the market, updated for Excel 2010 As the world's leading spreadsheet application, Excel has a huge user base. The release of Office 2010 brings major changes to Excel, so Excel For Dummies comes to the rescue once more! In the friendly and non-threatening For Dummies style, this popular guide shows beginners how to get up and running with Excel and helps more experienced users get comfortable with new features.Excel is the number one spreadsheet application worldwide, and Excel For Dummies is the number one guide to using i

  12. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  13. Shutdown-induced tensile stress in monolithic miniplates as a possible cause of plate pillowing at very high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Medvedev, Pavel G [Idaho National Laboratory; Ozaltun, Hakan [Idaho National Laboratory; Robinson, Adam Brady [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory

    2014-04-01

    Post-irradiation examination of Reduced Enrichment for Research and Test Reactors (RERTR)-12 miniplates showed that in-reactor pillowing occurred in at least 4 plates, rendering performance of these plates unacceptable. To address in-reactor failures, efforts are underway to define the mechanisms responsible for in-reactor pillowing, and to suggest improvements to the fuel plate design and operational conditions. To achieve these objectives, the mechanical response of monolithic fuel to fission and thermally-induced stresses was modeled using a commercial finite element analysis code. Calculations of stresses and deformations in monolithic miniplates during irradiation and after the shutdown revealed that the tensile stress generated in the fuel increased from 2 MPa to 100 MPa at shutdown. The increase in tensile stress at shutdown possibly explains in-reactor pillowing of several RERTR-12 miniplates irradiated to the peak local burnup of up to 1.11x1022 fissions/cm3 . This paper presents the modeling approach and calculation results, and compares results with post-irradiation examinations and mechanical testing of irradiated fuel. The implications for the safe use of the monolithic fuel in research reactors are discussed, including the influence of fuel burnup and power on the magnitude of the shutdown-induced tensile stress.

  14. Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

  15. The burnup dependence of light water reactor spent fuel oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

  16. Excellent Teachers' Perspectives on Excellent Teaching

    Science.gov (United States)

    Keeley, Jared W.; Ismail, Emad; Buskist, William

    2016-01-01

    Studies of master teaching have investigated a set of qualities that define excellent teaching. However, few studies have investigated master teachers' perspectives on excellent teaching and how it may differ from other faculty or students. The current study investigated award-winning teachers' (N = 50) ratings of the 28 qualities on the teacher…

  17. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  18. Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems

    Directory of Open Access Journals (Sweden)

    Thomas Frosio

    2017-01-01

    Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.

  19. Burnup measurements on spent fuel elements of the RP-10 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vela Mora, Mariano; Gallardo Padilla, Alberto; Palomino, Jose Luis Castro, E-mail: mvela@ipen.gob.p [Instituto Peruano de Energia Nuclear (IPEN/Peru), Lima (Peru). Grupo de Calculo, Analisis y Seguridad de Reactores; Terremoto, Luis Antonio Albiac, E-mail: laaterre@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using {sup 137}Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%. (author)

  20. Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations

    Science.gov (United States)

    Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.

    2014-04-01

    The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.

  1. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [ORNL; Mueller, Don [ORNL; Wagner, John C [ORNL

    2011-01-01

    One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission

  2. Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Parks, Cecil V [ORNL; Wagner, John C [ORNL; Mueller, Don [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    In the United States (U.S.) there has been and continues to be considerable interest in the increased use of burnup credit as part of the safety basis for SNF systems and this interest has motivated numerous technical studies related to the application of burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission initiated a burnup credit research program, with support from the Oak Ridge National Laboratory, to develop regulatory guidance and the supporting technical basis for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details.

  3. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  4. Application of EPMA data for the development of the code systems TRANSURANUS and ALEPH.

    Science.gov (United States)

    Haeck, Wim; Verboomen, Bernard; Schubert, Arndt; Van Uffelen, Paul

    2007-06-01

    In the present article, electron probe microanalysis data for Pu and Nd is being used for validating the predictions of the radial power profile in a nuclear fuel rod at an ultrahigh burn-up of 95 and 102 MWd/kgHM. As such the validation of both the new Monte Carlo burn-up code ALEPH and the simpler TUBRNP model of the fuel rod performance code TRANSURANUS has been extended. The analysis of the absolute concentrations and individual isotopes also indicates potential improvements in the predictive capabilities of the simple TUBRNP model, based on the one-group cross sections inferred from the neutron transport calculations in the ALEPH code. This is a first important step toward extending the application range of the fuel rod performance code to burn-up values projected in nuclear power rods based on current trends.

  5. A Criticality Evaluation of the GBC-32 Dry Storage Cask in PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyoungju; Park, Kwangheon; Hong, Ser Gi [Kyung Hee Univ., Yongin (Korea, Republic of)

    2015-05-15

    The current criticality safety evaluation assumes the only unirradiated fresh fuels with the maximum enrichment in a dry storage cask (DSC) for conservatism without consideration of the depletion of fissile nuclides and the generation of neutron-absorbing fission products. However, the large conservatism leads to the significant increase of the storage casks required. Thus, the application of burnup credit which takes credit for the reduction of reactivity resulted from fuel depletion can increase the capacity in storage casks. On the other hand, the burnup credit application introduces lots of complexity into a criticality safety analysis such as the accurate estimation of the isotopic inventories and the burnup of UNFs and the validation of the criticality calculation. The criticality evaluation with an effect of burnup credit was performed for the DSC of GBC-32 by using SCALE 6.1/STARBUCS. keff values were calculated as a function of burnup and cooling time for four initial enrichments of 2, 3, 4, and 5 wt. % 235U. The values were calculated for the burnup range of 0 to 60,000 MWD/MTU, in increments of 10,000 MWD/MTU, and for five cooling times of 0, 5, 10, 20, and 40 years.

  6. Master VISUALLY Excel 2010

    CERN Document Server

    Marmel, Elaine

    2010-01-01

    The complete visual reference on Excel basics. Aimed at visual learners who are seeking an all-in-one reference that provides in-depth coveage of Excel from a visual viewpoint, this resource delves into all the newest features of Excel 2010. You'll explore Excel with helpful step-by-step instructions that show you, rather than tell you, how to navigate Excel, work with PivotTables and PivotCharts, use macros to streamline work, and collaborate with other users in one document.: This two-color guide features screen shots with specific, numbered instructions so you can learn the actions you need

  7. Excel Dashboards & Reports

    CERN Document Server

    Alexander, Michael

    2010-01-01

    The go to resource for how to use Excel dashboards and reports to better conceptualize data. Many Excel books do an adequate job of discussing the individual functions and tools that can be used to create an "Excel Report." What they don't offer is the most effective ways to present and report data. Offering a comprehensive review of a wide array of technical and analytical concepts, Excel Reports and Dashboards helps Excel users go from reporting data with simple tables full of dull numbers, to presenting key information through the use of high-impact, meaningful reports and dashboards that w

  8. Excel 2010 Made Simple

    CERN Document Server

    Katz, Abbott

    2011-01-01

    Get the most out of Excel 2010 with Excel 2010 Made Simple - learn the key features, understand what's new, and utilize dozens of time-saving tips and tricks to get your job done. Over 500 screen visuals and clear-cut instructions guide you through the features of Excel 2010, from formulas and charts to navigating around a worksheet and understanding Visual Basic for Applications (VBA) and macros. Excel 2010 Made Simple takes a practical and highly effective approach to using Excel 2010, showing you the best way to complete your most common spreadsheet tasks. You'll learn how to input, format,

  9. Excel 2013 formulas

    CERN Document Server

    Walkenbach, John

    2013-01-01

    Maximize the power of Excel 2013 formulas with this must-have Excel reference John Walkenbach, known as ""Mr. Spreadsheet,"" is a master at deciphering complex technical topics and Excel formulas are no exception. This fully updated book delivers more than 800 pages of Excel 2013 tips, tricks, and techniques for creating formulas that calculate, developing custom worksheet functions with VBA, debugging formulas, and much more. Demonstrates how to use all the latest features in Excel 2013 Shows how to create financial formulas and tap into the power of array formulas

  10. Excel 2013 simplified

    CERN Document Server

    McFedries, Paul

    2013-01-01

    A friendly, visual approach to learning the basics of Excel 2013 As the world's leading spreadsheet program, Excel is a spreadsheet and data analysis tool that is part of the Microsoft Office suite. The new Excel 2013 includes new features and functionalities that require users of older versions to re-learn the application. However, whether you're switching from an earlier version or learning Excel for the first time, this easy-to-follow visual guide gets you going with Excel 2013 quickly and easily. Numbered steps as well as full-color screen shots, concise information, and helpfu

  11. Clinical excellence in cardiology.

    Science.gov (United States)

    Ziegelstein, Roy C

    2011-08-15

    A recent study identified 7 domains of clinical excellence on the basis of interviews with "clinically excellent" physicians at academic institutions in the United States: (1) communication and interpersonal skills, (2) professionalism and humanism, (3) diagnostic acumen, (4) skillful negotiation of the health care system, (5) knowledge, (6) taking a scholarly approach to clinical practice, and (7) having passion for clinical medicine. What constitutes clinical excellence in cardiology has not previously been defined. The author discusses clinical excellence in cardiology using the framework of these 7 domains and also considers the additional domain of clinical experience. Specific aspects of the domains of clinical excellence that are of greatest relevance to cardiology are highlighted. In conclusion, this discussion characterizes what constitutes clinical excellence in cardiology and should stimulate additional discussion of the topic and an examination of how the domains of clinical excellence in cardiology are related to specific patient outcomes. Copyright © 2011 Elsevier Inc. All rights reserved.

  12. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  13. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  14. Excel 2010 bible

    CERN Document Server

    Walkenbach, John

    2010-01-01

    A comprehensive reference to the newest version of the world's most popular spreadsheet application: Excel 2010 John Walkenbach's name is synonymous with excellence in computer books that decipher complex technical topics. Known as ""Mr. Spreadsheet,"" Walkenbach shows you how to maximize the power of all the new features of Excel 2010. An authoritative reference, this perennial bestseller proves itself indispensable no matter your level of skill, from Excel beginners and intermediate users to power users and potential power users everywhere. Fully updated for the new release, this

  15. A Simple Formula for Local Burnup and Isotope Distributions Based on Approximately Constant Relative Reaction Rate

    Directory of Open Access Journals (Sweden)

    Cenxi Yuan

    2016-01-01

    Full Text Available A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of 235U, 238U, and 239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except (n,γ of  238U and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption.

  16. Excel dashboards and reports

    CERN Document Server

    Alexander, Michael

    2013-01-01

    Learn to use Excel dashboards and reports to better conceptualize data Updated for all the?latest features and capabilities of Excel 2013, this go-to resource provides you with in-depth coverage of the individual functions and tools that can be used to create?compelling Excel reports. Veteran author Michael Alexander walks you through the most effective ways to present and report data. Featuring a comprehensive review of a wide array of technical and analytical concepts, this essential guide helps you go from reporting data with simple tables full of dull numbers to presenting

  17. Excel for Cost Engineers

    Science.gov (United States)

    Butts, Glenn C.

    2007-01-01

    Excel is a powerful tool with a plethora of largely unused capabilities that can make the life of an engineer cognizant of them a great deal easier. This paper offers tips, tricks and techniques for better worksheets. Including the use of data validation, conditional formatting, subtotals, text formulas, custom functions and much more. It is assumed that the reader will have a cursory understanding of Excel so the basics will not be covered, if you get hung up try Excel's built in help menus, or a good book.

  18. Excel2003 Formulas

    CERN Document Server

    Walkenbach, John

    2011-01-01

    Everything you need to know about* Mastering operators, error values, naming techniques, and absolute versus relative references* Debugging formulas and using the auditing tools* Importing and exporting XML files and mapping the data to specific cells* Using Excel 2003's rights management feature* Working magic with array formulas* Developing custom formulas to produce the results you needHere's the formula for Excel excellenceFormulas are the lifeblood of spreadsheets, and no one can bring a spreadsheet to life like John Walkenbach. In this detailed reference guide, he delves deeply into unde

  19. Excel 2003 for dummies

    CERN Document Server

    Harvey, Greg

    2013-01-01

    Every time you turn around, you run into Excel. It's on yourPC at work. It's on your PC at home. You get Excel files fromyour boss. Wouldn't you like to understand this powerfulMicrosoft Office spreadsheet program, once and for all? Now, youcan crunch financial data, add sparkle to presentations, convertstatic lists of numbers into impressive charts, and discover whatall the shouting's about regarding databases, formulas, andcells. You may even decide that getting organized with a goodspreadsheet is downright useful and fun! Flip open Excel 2003 For Dummies, and you'llquickly start getting th

  20. EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Shimskey, R. W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Klymyshyn, N. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, R. A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, P. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-15

    The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.

  1. An iterative approach for TRIGA fuel burn-up determination using nondestructive gamma-ray spectrometry.

    Science.gov (United States)

    Wang, T K; Peir, J J

    2000-01-01

    The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio.

  2. Sustainable Enterprise Excellence

    DEFF Research Database (Denmark)

    Edgeman, Rick

    2013-01-01

    Structured Abstract Purpose: Sustainable Enterprise Excellence (SEE) is defined and developed through integration and expansion of business excellence modeling and sustainability thought. The intent is to enable simple yet reliable enterprise assessment of triple bottom line (TBL) performance...... and produce actionable enterprise foresight that can enable next best practices and sources of sustainable competitive advantage through innovation. Methodology: Key elements of SEE are identified from various business excellence and sustainability reporting sources, including the Global Reporting Initiative...... assessment approach similar in structure to those behind established excellence awards are developed that enable enterprise assessment of progress toward SEE. The resulting assessment is delivered in a highly consumable, combined narrative and graphic format referred to as a SEE NEWS Report. Practical...

  3. Striving for Excellence

    DEFF Research Database (Denmark)

    Hattke, Fabian; Blaschke, Steffen

    2015-01-01

    no significant effects. Besides top management team composition, we find that a high number of faculties and a broad inclusion of internal status groups (students, tenured faculty, academic and administrative staff) and external stakeholders in decision making processes may enhance academic excellence......, the study contributes to the body of literature on strategic leadership in pluralistic organizations. It highlights the importance of heterogeneous governance structures and modular organization designs for achieving academic excellence. Practical implications: - The paper may inform practitioners...

  4. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  5. Study on core physics characteristics of high burn-up full MOX PWR core. 2

    Energy Technology Data Exchange (ETDEWEB)

    Kugo, Teruhiko; Okubo, Tsutomu; Shimada, Syoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    As one of options for future light water reactors, we have been studying a new concept of a high burn-up full MOX PWR core with a discharge burn-up of 100 GWd/t and a 3-year operation cycle being based on the existing light water reactor technology. We have already confirmed the feasibility of the core, in which a moderator to fuel volume ratio(Vm/Vf) is increased to 2.6 with the same fuel pin diameter of 9.5 mm as in the current PWR but with the enlarged fuel pin pitch of 13.8 mm. In this report, to improve the neutronics and thermal hydraulic performance of the high burn-up core, we subsequently propose a 600 MWe core ensuring discharge burn-up of 100 GWd/t by increasing Vm/Vf to 3.0 with the same fuel pin pitch of 12.6 mm as in the current PWR and the smaller fuel rod diameter of 8.3 mm instead of 9.5 mm. We have investigated its core characteristics in neutronics and confirmed its feasibility. The core neutronics performance is compared between Vm/Vf = 2.6 and 3.0. From the comparison, it is found that the proposed core with Vm/Vf 3.0 has more promising characteristics than with Vm/Vf = 2.6 such as saving of a fissile plutonium content of 0.3wt%, improvement in a departure from nucleate boiling ratio (DNBR) and so on, except for a shortened cycle length by 9%. In addition, we have investigated a low-leakage refueling scheme for both types of high burn-up cores. Without modification to fuel material such as addition of burnable poison and/or transuranium isotopes, it can not be expected to improve the burn-up efficiency by the low-leakage refueling scheme. (author)

  6. Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

    2016-11-01

    Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

  7. Nuclear Energy Research Initiative. Development of a Stabilized Light Water Reactor Fuel Matrix for Extended Burnup

    Energy Technology Data Exchange (ETDEWEB)

    BD Hanson; J Abrefah; SC Marschman; SG Prussin

    2000-09-08

    The main objective of this project is to develop an advanced fuel matrix capable of achieving extended burnup while improving safety margins and reliability for present operations. In the course of this project, the authors improve understanding of the mechanism for high burnup structure (HBS) formation and attempt to design a fuel to minimize its formation. The use of soluble dopants in the UO{sub 2} matrix to stabilize the matrix and minimize fuel-side corrosion of the cladding is the main focus.

  8. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishijima, Kiyomi; Fuketa, Toyoshi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  9. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  10. Textbooks for Responsible Data Analysis in Excel

    Science.gov (United States)

    Garrett, Nathan

    2015-01-01

    With 27 million users, Excel (Microsoft Corporation, Seattle, WA) is the most common business data analysis software. However, audits show that almost all complex spreadsheets have errors. The author examined textbooks to understand why responsible data analysis is taught. A purposeful sample of 10 textbooks was coded, and then compared against…

  11. The renewal of excellence.

    Science.gov (United States)

    Halawi, Mohamad J

    2014-10-01

    Excellence in orthopaedic surgery has been a long time in the making. What we currently enjoy reflects over a century of innovation, leadership, and consummate professionalism. Today, our practice faces monumental challenges that have never been more demanding, and our commitment to the preservation and advancement of excellence has never been more pressing. Away from the scientific tone but without romanticizing our proud heritage, this article seeks to revitalize the values that have defined our practice, drawing on the wisdom and experiences of leaders in our field. Collectively, it aims to stimulate our enlightened understanding, to reignite our hopes, and to reaffirm our commitment to the highest standards of excellence. Copyright © 2014 by The Journal of Bone and Joint Surgery, Incorporated.

  12. Paternity analysis in Excel.

    Science.gov (United States)

    Rocheta, Margarida; Dionísio, F Miguel; Fonseca, Luís; Pires, Ana M

    2007-12-01

    Paternity analysis using microsatellite information is a well-studied subject. These markers are ideal for parentage studies and fingerprinting, due to their high-discrimination power. This type of data is used to assign paternity, to compute the average selfing and outcrossing rates and to estimate the biparental inbreeding. There are several public domain programs that compute all this information from data. Most of the time, it is necessary to export data to some sort of format, feed it to the program and import the output to an Excel book for further processing. In this article we briefly describe a program referred from now on as Paternity Analysis in Excel (PAE), developed at IST and IBET (see the acknowledgments) that computes paternity candidates from data, and other information, from within Excel. In practice this means that the end user provides the data in an Excel sheet and, by pressing an appropriate button, obtains the results in another Excel sheet. For convenience PAE is divided into two modules. The first one is a filtering module that selects data from the sequencer and reorganizes it in a format appropriate to process paternity analysis, assuming certain conventions for the names of parents and offspring from the sequencer. The second module carries out the paternity analysis assuming that one parent is known. Both modules are written in Excel-VBA and can be obtained at the address (www.math.ist.utl.pt/~fmd/pa/pa.zip). They are free for non-commercial purposes and have been tested with different data and against different software (Cervus, FaMoz, and MLTR).

  13. Sustainable Enterprise Excellence

    DEFF Research Database (Denmark)

    Edgeman, Rick; Williams, Joseph; Eskildsen, Jacob Kjær

    Sustainable Enterprise Excellence balances complementary and competing interests of key stakeholder segments, including society and the natural environment and increases the likelihood of superior and sustainable competitive positioning and hence long-term enterprise success that is defined...... by continuously relevant and responsible governance, strategy, actions and performance consistent with high-level organizational resilience, robustness and resplendence (R3). This is accomplished through organizational design and function emphasizing innovation, enterprise intelligence & analytics, operational......, supply chain, customer-related, human capital, financial, marketplace, societal, and environmental performance. Sustainable Enterprise Excellence integrates ethical, efficient and effective (E3) enterprise governance with 3E (equity, ecology, economy) Triple Top Line strategy throughout enterprise...

  14. Microsoft Excel 2007 - Manual

    OpenAIRE

    Roque, Vítor

    2009-01-01

    O Microsoft Excel faz parte do "Office", um pacote de produtos que combinam vários tipos de software para criar documentos, folhas de cálculo, apresentações e para gerir correio electrónico. Nesta nova versão, o Microsoft Excel surge com uma nova interface, onde os tradicionais menus de opções desaparecem e dão lugar a um friso com diversos separadores onde podemos encontrar, mais facilmente, os comandos necessários à execução das tarefas que nos propomos realizar. O obje...

  15. Searching for excellence

    Science.gov (United States)

    Yager, Robert E.

    Visits to six school districts which were identified by the National Science Teachers Association's Search for Excellence program were made during 1983 by teams of 17 researchers. The reports were analyzed in search for common characteristics that can explain the requirements necessary for excellent science programs. The results indicate that creative ideas, administrative and community involvement, local ownership and pride, and well-developed in-service programs and implementation strategies are vital. Exceptional teachers with boundless energies also seem to exist where exemplary science programs are found.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  17. The Gd-isotopic fuel for high burnup in PWR's

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Marcio Soares; Mattos, João Roberto L. de; Andrade, Edison Pereira de, E-mail: marciod@cdtn.br, E-mail: jrmattos@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    Today, the discussion about the high burnup fuel is beyond the current fuel enrichment licensing and burnup limits. Licensing issues and material/design developments are again key features in further development of the LWR fuel design. Nevertheless, technological and economical solutions are already available or will be available in a short time. In order to prevent the growth of the technological gap, Brazil's nuclear sector needs to invest in the training of new human resources, in the access to international databases, and in the upgrading existing infrastructure. Experimental database and R&D infrastructure are essential components to support the autonomous development of Brazilian Nuclear Reactors, promoting the development of national technologies. The (U,Gd)O{sub 2} isotopic fuel proposed by the CDTN's staff solve two main issues in the high burnup fuel, which are (1) the peak of reactivity resulting from the Gd-157 fast burnup, and (2) the peak of temperature in the (U,Gd)O{sub 2} nuclear fuel resulting from detrimental effects in the thermal properties for gadolinia additions higher than 2%. A sustainable future can be envisaged for the nuclear energy. (author)

  18. About a fuel for burnup reactor of periodical pulsed nuclear pumped laser

    Energy Technology Data Exchange (ETDEWEB)

    Volkov, A.I.; Lukin, A.V.; Magda, L.E.; Magda, E.P.; Pogrebov, I.S.; Putnikov, I.S.; Khmelnitsky, D.V.; Scherbakov, A.P. [Russian Federal Nuclear Center, Snezhinsk (Russian Federation)

    1998-07-01

    A physical scheme of burnup reactor for a Periodic Pulsed Nuclear Pumped Laser was supposed. Calculations of its neutron physical parameters were made. The general layout and construction of basic elements of the reactor are discussed. The requirements for the fuel and fuel elements are established. (author)

  19. Modeling of Pore Coarsening in the Rim Region of High Burn-up UO2 Fuel

    Directory of Open Access Journals (Sweden)

    Hongxing Xiao

    2016-08-01

    Full Text Available An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS of irradiated UO2 fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO2 fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO2 fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

  20. Dependence of heavy metal burnup on nuclear data libraries for fast reactors

    CERN Document Server

    Ohki, S

    2003-01-01

    Japan Nuclear Cycle Development Institute (JNC) is considering the highly burnt fuel as well as the recycling of minor actinide (MA) in the development of commercialized fast reactor cycle systems. Higher accuracy in burnup calculation is going to be required for higher mass plutonium isotopes ( sup 2 sup 4 sup 0 Pu, etc.) and MA nuclides. In the framework of research and development aiming at the validation and necessary improvements of fast reactor burnup calculation, we investigated the differences among the burnup calculation results with the major nuclear data libraries: JEF-2.2, ENDF/B-VI Release 5, JENDL-3.2, and JENDL-3.3. We focused on the heavy metal nuclides such as plutonium and MA in the central core region of a conventional sodium-cooled fast reactor. For main heavy metal nuclides ( sup 2 sup 3 sup 5 U, sup 2 sup 3 sup 8 U, sup 2 sup 3 sup 9 Pu, sup 2 sup 4 sup 0 Pu, and sup 2 sup 4 sup 1 Pu), number densities after 1-cycle burnup did not change over one or two percent. Library dependence was re...

  1. A complete NUHOMS {sup registered} solution for storage and transport of high burnup spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bondre, J. [Transnuclear, Inc. (AREVA Group), Fremont, CA (United States)

    2004-07-01

    The discharge burnups of spent fuel from nuclear power plants keep increasing with plants discharging or planning to discharge fuel with burnups in excess of 60,000 MWD/MTU. Due to limited capacity of spent fuel pools, transfer of older cooler spent fuel from fuel pool to dry storage, and very limited options for transport of spent fuel, there is a critical need for dry storage of high burnup, higher heat load spent fuel so that plants could maintain their full core offload reserve capability. A typical NUHOMS {sup registered} solution for dry spent fuel storage is shown in the Figure 1. Transnuclear, Inc. offers two advanced NUHOMS {sup registered} solutions for the storage and transportation of high burnup fuel. One includes the NUHOMS {sup registered} 24PTH system for plants with 90.7 Metric Ton (MT) crane capacity; the other offers the higher capacity NUHOMS {sup registered} 32PTH system for higher crane capacity. These systems include NUHOMS {sup registered} - 24PTH and -32PTH Transportable Canisters stored in a concrete storage overpack (HSM-H). These canisters are designed to meet all the requirements of both storage and transport regulations. They are designed to be transported off-site either directly from the spent fuel pool or from the storage overpack in a suitable transport cask.

  2. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  3. Coping with Excellence.

    Science.gov (United States)

    Culler, Jonathan

    2002-01-01

    Suggests that when the university devotes itself to "excellence," it substitutes, for a vision of what should be taught, a contentless measure, a bureaucratic concept that extends the range of managerial control without presuming to make judgments about content. Suggests that English departments should try to design sequences of courses that are…

  4. Statistische toetsen in Excel

    NARCIS (Netherlands)

    A.A.J. van Peet; H.A. Everaert

    2007-01-01

    In KG-18 hebben we verschillende grafieken laten zien. Een veel gebruikte grafische figuur in de statistiek is de zogenaamde boxplot. Weliswaar wordt deze niet vermeld bij de standaard keuzemogelijkheden onder de Wizard Grafieken, toch is het mogelijk om deze figuur met behulp van Excel te tekenen.

  5. Enlisting Excel--Again

    Science.gov (United States)

    Parramore, Keith

    2009-01-01

    In Volume 26, Number 2, we reported on a group case study run for level 3 mathematics students at the University of Brighton. At the core of the study was a quadratic assignment problem, and we reported on attempts by students to use Excel to solve the problem, and on the attendant difficulties. We provided an elegant solution. In this article, we…

  6. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jayalal, M.L., E-mail: jayalal@igcar.gov.in [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Ramachandran, Suja [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Rathakrishnan, S. [Reactor Physics Section, Madras Atomic Power Station (MAPS), Kalpakkam, Tamil Nadu (India); Satya Murty, S.A.V. [Electronics, Instrumentation and Radiological Safety Group (EIRSG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India); Sai Baba, M. [Resources Management Group (RMG), Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamil Nadu (India)

    2015-01-15

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  7. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  8. Analysis of burnup of Angra 2 PWR nuclear with addition of thorium dioxide fuel using ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    Goncalves, Isadora C.; Wichrowski, Caio C.; Oliveira, Claudio L. de; Vellozo, Sergio O.; Baptista, Camila O., E-mail: isadora.goncalves@ime.eb.br, E-mail: wichrowski@ime.eb.br, E-mail: d7luiz@yahoo.com.br, E-mail: vellozo@ime.eb.br, E-mail: camila.oliv.baptista@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear

    2017-11-01

    It is known that isotope {sup 232}thorium is a fertile nuclide with the ability to convert into {sup 233}uranium, a potentially fissile isotope, after absorbing a neutron. As there is a large stock of available thorium in the world, this element shows great promise in mitigate the world energy crisis, more particularly in the problem of uranium scarcity, besides being an alternative nuclear fuel for those currently used in reactors, and yet presenting advantages as an option for the non-proliferation movement, among others. In this study, the analysis of the remaining nuclides of burnup was carried out for the core configuration of a PWR (pressurized water reactor) reactor, specifically the Angra 2 reactor, using only uranium dioxide, its current configuration, and in different configurations including a mixed oxide of uranium and thorium in three concentrations, allowing a preliminary assessment of the feasibility of the modification of the fuel, the resulting production of {sup 233}uranium, the emergence of {sup 231}protactinium (an isotope that only occurs as a fission product of {sup 232}Th) resulting from burning. The study was carried out using data obtained from FSAR (Final Safety Analysis Report) of Angra 2, using the SCALE 6.1, a modeling and simulation nuclear code, especially its ORIGEN-ARP module, which analyzes the depletion of isotopes presents in a reactor. (author)

  9. A Concise Design for the Irradiation of U–10Zr Metallic Fuel at a Very Low Burnup

    Directory of Open Access Journals (Sweden)

    Haibing Guo

    2017-06-01

    Full Text Available In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

  10. A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang (China)

    2017-06-15

    In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

  11. Thermo-mechanical description of a nuclear pin, BACO code version 2.20

    Energy Technology Data Exchange (ETDEWEB)

    Marino, A.C. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Gerencia de Area Ciclo de Combustible; Savino, E.; Harriague, S. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Gerencia de Area Investigacion y Desarrollo

    1995-12-31

    BACO code, version 2.20 and some applications are presented. BACO (BArra COmbustible) is a code for the simulation of the thermo-mechanical and fission gas behavior of a cylindrical fuel rod under operation. The new version was developed in connection with the CRP FUMEX of the IAEA (Coordinated Research Project on Fuel Modelling at Extended Burnup). The project originated a conceptual revision of the original code. The revision includes convergence criteria, mathematical treatments and fuel behavior modelling. The code use domain is in PHWR fuel but it may be extended to other applications. The BACO code has a good performance in the range of low-intermediate burnup. We include the study of pore migration and restructuring, relocation of pellet fragments and gap heat conductance, fuel MOX rod analysis, and an example of FUMEX case. (author). 19 refs., 9 figs., 1 tab.

  12. Excel 2007 Charts

    CERN Document Server

    Walkenbach, John

    2008-01-01

    Excel, the top number-crunching tool, now offers a vastly improved charting function to help you give those numbers dimension and relativity. John Walkenbach, a.k.a. Mr. Spreadsheet, clearly explains all these charting features and shows you how to choose the right chart for your needs. You'll learn to modify data within the chart, deal with missing data, format your chart, use trend lines, construct "impossible" charts, create charts from pivot tables, dress them up with graphics, and more. Note: CD-ROM/DVD and other supplementary materials are not included as part of eBook file.

  13. Excel Programming Your Visual Blueprint for Creating Interactive Spreadsheets

    CERN Document Server

    Etheridge, Denise

    2010-01-01

    A great guide to Excel programming that is perfect for visual learners and takes you beyond Excel basics!. This book is the perfect reference for Excel users who want to delve deeper into the application to create powerful and dynamic programs. From creating macros to customizing dialog boxes, this step-by-step guide helps you get more out of Excel than you knew was possible. Each step has callouts so you can see exactly where the action takes place and this Web site offers tons of usable code and sample macros that you can put to use instantly.: Explains step-by-step how to automate Excel, th

  14. Excel 2010 Workbook for Dummies

    CERN Document Server

    Harvey, Greg

    2010-01-01

    Reinforce your understanding of Excel with these Workbook exercises. Boost your knowledge of important Excel tasks by putting your skills to work in real-world situations. The For Dummies Workbook format provides more than 100 exercises that help you create actual results with Excel so you can gain proficiency. Perfect for students, people learning Excel on their own, and financial professionals who must plan and execute complex projects in Excel, Excel 2010 Workbook For Dummies helps you discover all the ways this program can work for you.: Excel is the world's most popular number-crunching p

  15. 18 excel spreadsheets by species and year giving reproduction and growth data. One excel spreadsheet of herbicide treatment chemistry.

    Data.gov (United States)

    U.S. Environmental Protection Agency — Excel spreadsheets by species (4 letter code is abbreviation for genus and species used in study, year 2010 or 2011 is year data collected, SH indicates data for...

  16. Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

    Science.gov (United States)

    Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

    2005-05-01

    The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

  17. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  18. Analysis of simulated high burnup nuclear fuel by laser induced breakdown spectroscopy

    Science.gov (United States)

    Singh, Manjeet; Sarkar, Arnab; Banerjee, Joydipta; Bhagat, R. K.

    2017-06-01

    Advanced Heavy Water Reactor (AHWR) grade (Th-U)O2 fuel sample and Simulated High Burn-Up Nuclear Fuels (SIMFUEL) samples mimicking the 28 and 43 GWd/Te irradiated burn-up fuel were studied using laser-induced breakdown spectroscopy (LIBS) setup in a simulated hot-cell environment from a distance of > 1.5 m. Resolution of 60 emission lines of fission products was identified. Among them only a few emission lines were found to generate calibration curves. The study demonstrates the possibility to investigate impurities at concentrations around hundreds of ppm, rapidly at atmospheric pressure without any sample preparation. The results of Ba and Mo showed the advantage of LIBS analysis over traditional methods involving sample dissolution, which introduces possible elemental loss. Limits of detections (LOD) under Ar atmosphere shows significant improvement, which is shown to be due to the formation of stable plasma.

  19. Experience with incomplete control rod insertion in fuel with burnup exceeding approximately 40 GWD/MTU

    Energy Technology Data Exchange (ETDEWEB)

    Kee, E. [Houston Lighting & Power Co., Wadworth, TX (United States)

    1997-01-01

    Analysis and measurement experience with fuel assemblies having incomplete control rod insertion at burnups of approximately 40 GWD/MTU is presented. Control rod motion dynamics and simplified structural analyses are presented and compared to measurement data. Fuel assembly growth measurements taken with the plant Refueling Machine Z-Tape are described and presented. Bow measurements (including plug gauging) are described and potential improvements are suggested. The measurements described and analysis performed show that sufficient guide tube bow (either from creep or yield buckling) is present in some high burnup assemblies to stop the control rods before they reach their full limit of travel. Recommendations are made that, if implemented, could improve cost performance related to testing and analysis activities.

  20. EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

  1. Evaluation of Fission Product Critical Experiments and Associated Biases for Burnup Credit Validation

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Rearden, Bradley T [ORNL; Reed, Davis Allan [ORNL

    2010-01-01

    One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

  2. Evaluation of the Frequency for Gas Sampling for the High Burnup Confirmatory Data Project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations.

  3. Investigation of burnup credit allowance in the criticality safety evaluation of spent fuel casks

    Energy Technology Data Exchange (ETDEWEB)

    Lake, W.H. (USDOE, Washington, DC (USA)); Sanders, T.L. (Sandia National Labs., Albuquerque, NM (USA)); Parks, C.V. (Oak Ridge National Lab., TN (USA))

    1990-01-01

    This presentation discusses work in progress on criticality analysis verification for designs which take account of the burnup and age of transported fuel. The work includes verification of cross section data, correlation with experiments, proper extension of the methods into regimes not covered by experiments, establishing adequate reactivity margins, and complete documentation of the project. Recommendations for safe operational procedures are included, as well as a discussion of the economic and safety benefits of such designs.

  4. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  5. Criticality Uncertainty Analysis of Spent Fuel Transport Cask applying Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gang Ug; Park, Jae Ho; Kim, Do Hyun [Korea Nuclear Engineering and Service Corp, Daejeon (Korea, Republic of); Kim, Tae Man; Yoon, Jeong Hyun [Korea Radioactive Waste Management Corporation, Daejeon (Korea, Republic of)

    2011-09-15

    In general, conventional criticality analyses for spent fuel transport/dry storage systems have been performed based on assumption of fresh fuel concerning the potential uncertainties from number density calculation of Transuranic and Fission Products in spent fuel. However, because of economic loss due to the excessive criticality margin, recently the design of transport/dry storage systems with Burnup Credit(BUC) application has been actively developed. The uncertainties in criticality analyses on transport/storage systems with BUC technique show strong dependence upon initial enrichment and burnup rate, whereas those in the conventional criticality evaluation based on fresh fuel assumption do not show such a dependence. In this study, regulatory-required uncertainties of the criticality analyses for BK 26 Cask, which is conceptually designed spent fuel transport cask with BUC corresponding to the limiting circumstances on nuclear power plants in Korea, are evaluated as a function of initial enrichment and burnup rate. Results of this study will be used as basic data for spent fuel loading curve of BK 26 Cask.

  6. High burnup fuel onset conditions in dry storage. Prediction of EOL rod internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L.E.

    2015-07-01

    During dry storage, cladding resistance to failure can be affected by several degrading mechanisms like creep or hydrides radial reorientation. The driving force of these effects is the stress at which the cladding is submitted. The maximum stress in the cladding is determined by the end-of-reactor-life (EOL) rod internal pressure, PEOL, at the maximum temperature attained during dry storage. Thus, PEOL sets the initial conditions of storage for potential time-dependent changes in the cladding. Based on FRAPCON-3.5 calculations, the aim of this work is to analyse the PEOL of a PWR fuel rod irradiated to burnups greater than 60 GWd/tU, where limited information is available. In order to be conservative, demanding irradiation histories have been used with a peak linear power of 44 kW/m. FRAPCON-3.5 results show an increasing exponential trend of PEOL with burnup, from which a simple correlation has been derived. The comparison with experimental data found in the literature confirms the enveloping nature of the predicted curve. Based on that, a conservative prediction of cladding stress in dry storage has been obtained. The comparison with a critical stress threshold related to hydrides embrittlement seems to point out that this issue should not be a concern at burnups below 65 GWd/tU. (Author)

  7. Preparation of higher-actinide burnup and cross section samples. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.

    1981-01-01

    A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were /sup 241/Am and /sup 244/Cm in the forms of Am/sub 2/O/sub 3/, Cm/sub 2/O/sub 3/, and Am/sub 6/ Cm(RE)/sub 7/O/sub 21/, where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program.

  8. Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

    Directory of Open Access Journals (Sweden)

    Lecarpentier D.

    2013-03-01

    Full Text Available Burnup Credit (BUC is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a “best estimate” value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library. Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

  9. Study on nuclear physics of high burnup full MOX PWR core

    Energy Technology Data Exchange (ETDEWEB)

    Kugo, Teruhiko; Shimada, Syoichiro; Okubo, Tsutomu; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    As one of options for future light water reactors, we have been studying a new concept of a high burnup full MOX PWR core. We have proposed a core of 600 MWe to ensure discharged burnup of 100 GWd/t by increasing a moderator to fuel volume ratio to 2.6 with enlarged fuel pin pitch of 13.8 mm, and have investigated its feasibility in neutronics. A plutonium fissile content of 12% is needed for this core. A soluble boric acid with B-10 enrichment of 40% is able to control burnup reactivity without increasing a capacity of boron tanks. A control rod cluster with use of natural boron carbide (B{sub 4}C) per three fuel assemblies ensures a shutdown margin of more than 2%dk/kk`. A moderator temperature and void coefficients are negative through an operating cycle. Although the use of burnable poisons like Gd{sub 2}O{sub 3} and Er{sub 2}O{sub 3} is not necessary to reduce an excess reactivity, it can lower a radial power peaking factor by about 0.1. (author)

  10. High burnup performance of Mg, Mg-Nb and Ti doped UO{sub 2} fuels

    Energy Technology Data Exchange (ETDEWEB)

    Shiratori, Tetsuo; Serizawa, Hiroyuki; Fukuda, Kousaku [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Fujino, Takeo; Sato, Nobuaki; Yamada, Kohta [Institute for Advanced Materials Processing, Tohoku Univ., Sendai, Miyagi (Japan)

    2000-09-01

    In order to control irradiation performance of fuel swelling and FP gas release etc. at high burnups of light water reactor fuels, doped UO{sub 2} pellet fuels were prepared and their irradiation behavior was examined. The UO{sub 2} pellets doped 2.5 to 15mol%Mg, 5mol%Mg - 5mol%Nb, and 3.5mol%Ti and undoped UO{sub 2} pellets as a reference fuel were loaded together in a capsule and irradiated to the maximum burnups of 94GWd/t(U) below temperature of 1000degC in the JRR-3M reactor of JAERI. As results of post-irradiation examinations such as visual inspection, dimensional and density change measurements, thermal diffusivity and ceramography with optical microscope and EPMA, no difference was observed between the doped and the reference UO{sub 2} fuels. And valuable results were obtained on high burnup properties for swelling rates, thermal conductivities, structure changes and so on. (author)

  11. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Venkiteswaran, C.N., E-mail: cnv@igcar.gov.in; Jayaraj, V.V.; Ojha, B.K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B.P.C.; Kasiviswanathan, K.V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel–clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel–clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  12. OECD/NEA burnup credit criticality benchmark. Result of phase IIA

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Makoto; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-02-01

    The report describes the final result of the Phase IIA of the Burnup Credit Criticality Benchmark conducted by OECD/NEA. In the Phase IIA benchmark problems, the effect of an axial burnup profile of PWR spent fuels on criticality (end effect) has been studied. The axial profiles at 10, 30 and 50 GWd/t burnup have been considered. In total, 22 results from 18 institutes of 10 countries have been submitted. The calculated multiplication factors from the participants have lain within the band of {+-} 1% {Delta}k. For the irradiation up to 30 GWd/t, the end effect has been found to be less than 1.0% {Delta}k. But, for the 50 GWd/t case, the effect is more than 4.0% {Delta}k when both actinides and FPs are taken into account, whereas it remains less than 1.0% {Delta}k when only actinides are considered. The fission density data have indicated the importance end regions have in the criticality safety analysis of spent fuel systems. (author).

  13. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  14. Preliminary 3D burn-up analysis of the HPLWR core

    Energy Technology Data Exchange (ETDEWEB)

    Monti, Lanfranco; Gabrielli, Fabrizio; Schulenberg, Thomas [Forschungszentrum Karlsruhe (Germany). Inst. for Nuclear and Energy Technologies

    2009-07-01

    The High Performance Light Water Reactor (HPLWR) is an innovative reactor concept cooled and moderated with water at supercritical pressure (25 MPa) whose feasibility is analyzed within a European framework [1]. The pronounced variation in water density, which takes place inside the core, is due to the coolant heat up from 550 K to 800 K and is supposed to generate pronounced 3D effects during reactor operation because the different core regions have different flux amplitude and neutron spectrum. Open questions are how k{sub eff} and the power-map will change during the burn-up and require a 3D multi-zone burn-up analysis of the core. This goal is achieved using the ERANOS system [2, 3], which is a deterministic tool for neutronic core analyses. The starting condition is taken from a neutronic/thermal-hydraulic coupled solution of the whole core [4], which does not yet include any fuel enrichment optimization nor reactivity control systems, i.e. control rods or burnable poisons. Uranium dioxide enriched to 5wt% in {sup 235}U is used as starting fuel while typical LWRs evolution chains for actinides and fission products have been selected. The core nodalization used in the coupled system is also adopted for multi-zone burn-up analysis: there are 462 zones with different material composition, 21 in axial direction and 22 in the horizontal plane. A burn-up period of 200 days ({approx_equal}6400 MWd/tHM) is considered here and has been divided into two different smaller time steps: 1) an inner time step at which macroscopic cross-sections (XSs) and the flux normalization are calculated according to the change in fuel isotopic composition; 2) an outer time step at which whole core flux calculations are performed to evaluate the region-wise neutron flux distribution. The length of the flux calculation time step has to be short enough to avoid unphysical power-shape oscillations, as underlined by Reiss et al. [5] with a different computational approach. The 40 groups

  15. SRAC95; general purpose neutronics code system

    Energy Technology Data Exchange (ETDEWEB)

    Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-03-01

    SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).

  16. Compound effects of operating parameters on burnup credit criticality analysis in boiling water reactor spent fuel assemblies

    Directory of Open Access Journals (Sweden)

    Shang-Chien Wu

    2018-02-01

    Full Text Available This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor (keff versus burnup (B are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE14 10 × 10 boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC-68 storage cask. The results revealed that the curves of keff versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of keff,Δk in some compound effects was not a summation of the all Δk resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of keff versus B for both single and compound effects.

  17. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  18. Criticality Analysis of Assembly Misload in a PWR Burnup Credit Cask

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J. C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2008-01-31

    The Interim Staff Guidance on bumup credit (ISG-8) for spent fuel in storage and transportation casks, issued by the Nuclear Regulatory Commission's Spent Fuel Project Office, recommends a bumup measurement for each assembly to confirm the reactor record and compliance with the assembly bumup value used for loading acceptance. This recommendation is intended to prevent unauthorized loading (misloading) of assemblies due to inaccuracies in reactor burnup records and/or improper assembly identification, thereby ensuring that the appropriate subcritical margin is maintained. This report presents a computational criticality safety analysis of the consequences of misloading fuel assemblies in a highcapacity cask that relies on burnup credit for criticality safety. The purpose of this report is to provide a quantitative understanding of the effects of fuel misloading events on safety margins. A wide variety of fuel-misloading configurations are investigated and results are provided for informational purposes. This report does not address the likelihood of occurrence for any of the misload configurations considered. For representative, qualified bumup-enrichment combinations, with and without fission products included, misloading two assemblies that are underburned by 75% results in an increase in keff of 0.025-0.045, while misloading four assemblies that are underburned by 50% also results in an increase in keff of 0.025-0.045. For the cask and conditions considered, a reduction in bumup of 20% in all assemblies results in an increase in kff of less than 0.035. Misloading a single fresh assembly with 3, 4, or 5 wt% 235U enrichment results in an increase in keffof--0.02, 0.04, or 0.06, respectively. The report concludes with a summary of these and other important findings, as well as a discussion of relevant issues that should be considered when assessing the appropriate role of burnup measurements.

  19. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  20. Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

    Science.gov (United States)

    Cozzo, C.; Staicu, D.; Pagliosa, G.; Papaioannou, D.; Rondinella, V. V.; Konings, R. J. M.; Walker, C. T.; Barker, M. A.; Hervé, P.

    2010-05-01

    The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO 2 LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded "laser-flash" device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO 2 of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.

  1. Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

    Energy Technology Data Exchange (ETDEWEB)

    Cozzo, C. [European Commission, Joint Research Centre, Institute for Transuranium Element, P.O. Box 2340, D-76125 Karlsruhe (Germany); Staicu, D., E-mail: dragos.staicu@ec.europa.e [European Commission, Joint Research Centre, Institute for Transuranium Element, P.O. Box 2340, D-76125 Karlsruhe (Germany); Pagliosa, G.; Papaioannou, D.; Rondinella, V.V.; Konings, R.J.M.; Walker, C.T. [European Commission, Joint Research Centre, Institute for Transuranium Element, P.O. Box 2340, D-76125 Karlsruhe (Germany); Barker, M.A. [The UK' s National Nuclear Laboratory Ltd., Central Laboratory, Sellafield, Seascale, Cumbria CA20 1PG (United Kingdom); Herve, P. [LEEE, University Paris X, 1 Chemin Desvalliere, 92410 Ville d' Avray (France)

    2010-05-31

    The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO{sub 2} LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded 'laser-flash' device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO{sub 2} of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.

  2. Excel VBA 24-hour trainer

    CERN Document Server

    Urtis, Tom

    2015-01-01

    Master VBA automation quickly and easily to get more out of Excel Excel VBA 24-Hour Trainer, 2nd Edition is the quick-start guide to getting more out of Excel, using Visual Basic for Applications. This unique book/video package has been updated with fifteen new advanced video lessons, providing a total of eleven hours of video training and 45 total lessons to teach you the basics and beyond. This self-paced tutorial explains Excel VBA from the ground up, demonstrating with each advancing lesson how you can increase your productivity. Clear, concise, step-by-step instructions are combined wit

  3. Excel data analysis for dummies

    CERN Document Server

    Nelson, Stephen L

    2014-01-01

    Harness the power of Excel to discover what your numbers are hiding Excel Data Analysis For Dummies, 2nd Edition is the ultimate guide to getting the most out of your data. Veteran Dummies author Stephen L. Nelson guides you through the basic and not-so-basic features of Excel to help you discover the gems hidden in your rough data. From input, to analysis, to visualization, the book walks you through the steps that lead to superior data analysis. Excel is the number-one spreadsheet application, with ever-expanding capabilities. If you're only using it to balance the books, you're missing out

  4. His Excellency Mr Farukh Amil

    CERN Multimedia

    hugho, Chemli

    2017-01-01

    His Excellency Mr Farukh Amil Ambassador Extraordinary and Plenipotentiary Permanent Representative of the Islamic Republic of Pakistan to the United Nations Office and other international organisations in Geneva

  5. Teaching physics using Microsoft Excel

    Science.gov (United States)

    Uddin, Zaheer; Ahsanuddin, Muhammad; Khan, Danish Ahmed

    2017-09-01

    Excel is both ubiquitous and easily understandable. Most people from every walk of life know how to use MS office and Excel spreadsheets. Students are also familiar with spreadsheets. Most students know how to use spreadsheets for data analysis. Besides basic use of Excel, some important aspects of spreadsheets are highlighted in this article. MS Excel can be used to visualize effects of various parameters in a physical system. It can be used as a simulating tool; simulation of wind data has been done through spreadsheets in this study. Examples of Lissajous figures and a damped harmonic oscillator are presented in this article.

  6. Learn Excel 2011 for Mac

    CERN Document Server

    Hart-Davis, Guy

    2011-01-01

    Microsoft Excel 2011 for Mac OS X is a powerful application, but many of its most impressive features can be difficult to find. Learn Excel 2011 for Mac by Guy Hart-Davis is a practical, hands-on approach to learning all of the details of Excel 2011 in order to get work done efficiently on Mac OS X. From using formulas and functions to creating databases, from analyzing data to automating tasks, you'll learn everything you need to know to put this powerful application to use for a variety of tasks. What you'll learn * The secrets of the Excel for Mac interface! * How to create effective workbo

  7. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  8. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Arayones, J.M.

    1985-06-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions.

  9. Used Fuel Disposition Campaign - Baseline Studies for Ring Compression Testing of High-Burnup Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Y. Y. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-11-23

    Structural analyses of high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior during long-term dry cask storage and transportation. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). On the basis of previous test results, susceptibility to radial-hydride precipitation depends on cladding material, microstructure, and pre-drying distribution of hydrides across the cladding wall, as well as peak hoop stresses and temperatures during drying operations and storage. Susceptibility to embrittlement depends on the extent of radial-hydride precipitation and the thickness of the outer-surface hydride rim. These results highlight the importance of determining the DBTT for high-burnup cladding as a function of peak drying-storage temperatures and stresses and including the relevant mechanical properties in cask structural analyses. Additional testing is needed at lower (and perhaps more realistic) peak drying-storage temperatures and stresses, for which the DBTT is expected to decrease.

  10. ARS-Media for Excel

    Science.gov (United States)

    ARS-Media for Excel is an ion solution calculator that uses Microsoft Excel to generate recipes of salts for complex ion mixtures specified by the user. Generating salt combinations (recipes) that result in pre-specified target ion values is a linear programming problem. Thus, the recipes are genera...

  11. Assessment of Fission Product Cross-Section Data for Burnup Credit Applications

    Energy Technology Data Exchange (ETDEWEB)

    Leal, Luiz C [ORNL; Derrien, Herve [ORNL; Dunn, Michael E [ORNL; Mueller, Don [ORNL

    2007-12-01

    Past efforts by the Department of Energy (DOE), the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), and others have provided sufficient technical information to enable the NRC to issue regulatory guidance for implementation of pressurized-water reactor (PWR) burnup credit; however, consideration of only the reactivity change due to the major actinides is recommended in the guidance. Moreover, DOE, NRC, and EPRI have noted the need for additional scientific and technical data to justify expanding PWR burnup credit to include fission product (FP) nuclides and enable burnup credit implementation for boiling-water reactor (BWR) spent nuclear fuel (SNF). The criticality safety assessment needed for burnup credit applications will utilize computational analyses of packages containing SNF with FP nuclides. Over the years, significant efforts have been devoted to the nuclear data evaluation of major isotopes pertinent to reactor applications (i.e., uranium, plutonium, etc.); however, efforts to evaluate FP cross-section data in the resonance region have been less thorough relative to actinide data. In particular, resonance region cross-section measurements with corresponding R-matrix resonance analyses have not been performed for FP nuclides. Therefore, the objective of this work is to assess the status and performance of existing FP cross-section and cross-section uncertainty data in the resonance region for use in burnup credit analyses. Recommendations for new cross-section measurements and/or evaluations are made based on the data assessment. The assessment focuses on seven primary FP isotopes (103Rh, 133Cs, 143Nd, 149Sm, 151Sm, 152Sm, and 155Gd) that impact reactivity analyses of transportation packages and two FP isotopes (153Eu and 155Eu) that impact prediction of 155Gd concentrations. Much of the assessment work was completed in 2005, and the assessment focused on the latest FP cross-section evaluations available in the

  12. EMG data warehouse with excel-based visual analytics

    OpenAIRE

    Lampret , Luka

    2012-01-01

    The aim of the thesis was to demonstrate that a state-of-the-art solution for management and analytical visualization of data can be implemented based on Excel 2010. We used a large and complex database of measurements of electrical activity in uterine smooth muscles during pregnancy in an animal model (pregnant sheep). The thesis comprises two parts. The first part fully exploits the functionalities of the latest version of Excel – ranging from advanced formulas through controls and VBA code...

  13. Teach yourself visually complete Excel

    CERN Document Server

    McFedries, Paul

    2013-01-01

    Get the basics of Excel and then go beyond with this new instructional visual guide While many users need Excel just to create simple worksheets, many businesses and professionals rely on the advanced features of Excel to handle things like database creation and data analysis. Whatever project you have in mind, this visual guide takes you step by step through what each step should look like. Veteran author Paul McFedries first presents the basics and then gradually takes it further with his coverage of designing worksheets, collaborating between worksheets, working with visual data

  14. Modeling of pore coarsening in the rim region of high burn-up UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Hongxing; Long, Chong Sheng [Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu (China)

    2016-08-15

    An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated UO{sub 2} fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO{sub 2} fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO{sub 2} fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

  15. Critical assessment of the pore size distribution in the rim region of high burnup UO{sub 2} fuels

    Energy Technology Data Exchange (ETDEWEB)

    Cappia, F. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Department of Nuclear Engineering, Faculty of Mechanical Engineering, Technische Universität München, D-85748 Garching bei München (Germany); Pizzocri, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Nuclear Engineering Division, Energy Department, Politecnico di Milano, 20156 Milano (Italy); Schubert, A.; Van Uffelen, P.; Paperini, G.; Pellottiero, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Macián-Juan, R. [Department of Nuclear Engineering, Faculty of Mechanical Engineering, Technische Universität München, D-85748 Garching bei München (Germany); Rondinella, V.V., E-mail: Vincenzo.RONDINELLA@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany)

    2016-11-15

    A new methodology is introduced to analyse porosity data in the high burnup structure. Image analysis is coupled with the adaptive kernel density estimator to obtain a detailed characterisation of the pore size distribution, without a-priori assumption on the functional form of the distribution. Subsequently, stereological analysis is carried out. The method shows advantages compared to the classical approach based on the histogram in terms of detail in the description and accuracy within the experimental limits. Results are compared to the approximation of a log-normal distribution. In the investigated local burnup range (80–200 GWd/tHM), the agreement of the two approaches is satisfactory. From the obtained total pore density and mean pore diameter as a function of local burnup, pore coarsening is observed starting from ≈100 GWd/tHM, in agreement with a previous investigation. - Highlights: • A new methodology to analyse porosity is introduced. • The method shows advantages compared to the histogram. • Pore density and mean diameter data vs. burnup are presented. • Pore coarsening is observed starting from ≈100 GWd/tHM.

  16. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Science.gov (United States)

    2013-11-12

    ... Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research and Development Project (CDP) AGENCY: Fuel Cycle Technologies, Office of Nuclear Energy, Department of Energy... multiple entities while comments are resolved. Dated: November 5, 2013. Jay Jones, Office of Fuel Cycle...

  17. LICSS - a chemical spreadsheet in microsoft excel.

    Science.gov (United States)

    Lawson, Kevin R; Lawson, Jonty

    2012-02-02

    Representations of chemical datasets in spreadsheet format are important for ready data assimilation and manipulation. In addition to the normal spreadsheet facilities, chemical spreadsheets need to have visualisable chemical structures and data searchable by chemical as well as textual queries. Many such chemical spreadsheet tools are available, some operating in the familiar Microsoft Excel environment. However, within this group, the performance of Excel is often compromised, particularly in terms of the number of compounds which can usefully be stored on a sheet. LICSS is a lightweight chemical spreadsheet within Microsoft Excel for Windows. LICSS stores structures solely as Smiles strings. Chemical operations are carried out by calling Java code modules which use the CDK, JChemPaint and OPSIN libraries to provide cheminformatics functionality. Compounds in sheets or charts may be visualised (individually or en masse), and sheets may be searched by substructure or similarity. All the molecular descriptors available in CDK may be calculated for compounds (in batch or on-the-fly), and various cheminformatic operations such as fingerprint calculation, Sammon mapping, clustering and R group table creation may be carried out.We detail here the features of LICSS and how they are implemented. We also explain the design criteria, particularly in terms of potential corporate use, which led to this particular implementation. LICSS is an Excel-based chemical spreadsheet with a difference:• It can usefully be used on sheets containing hundreds of thousands of compounds; it doesn't compromise the normal performance of Microsoft Excel• It is designed to be installed and run in environments in which users do not have admin privileges; installation involves merely file copying, and sharing of LICSS sheets invokes automatic installation• It is free and extensibleLICSS is open source software and we hope sufficient detail is provided here to enable developers to add their

  18. Excel2Genie: A Microsoft Excel application to improve the flexibility of the Genie-2000 Spectroscopic software.

    Science.gov (United States)

    Forgács, Attila; Balkay, László; Trón, Lajos; Raics, Péter

    2014-12-01

    Excel2Genie, a simple and user-friendly Microsoft Excel interface, has been developed to the Genie-2000 Spectroscopic Software of Canberra Industries. This Excel application can directly control Canberra Multichannel Analyzer (MCA), process the acquired data and visualize them. Combination of Genie-2000 with Excel2Genie results in remarkably increased flexibility and a possibility to carry out repetitive data acquisitions even with changing parameters and more sophisticated analysis. The developed software package comprises three worksheets: display parameters and results of data acquisition, data analysis and mathematical operations carried out on the measured gamma spectra. At the same time it also allows control of these processes. Excel2Genie is freely available to assist gamma spectrum measurements and data evaluation by the interested Canberra users. With access to the Visual Basic Application (VBA) source code of this application users are enabled to modify the developed interface according to their intentions. Copyright © 2014 Elsevier Ltd. All rights reserved.

  19. ARS-Media for excel instruction manual

    Science.gov (United States)

    ARS-Media for Excel Instruction Manual is the instruction manual that explains how to use the Excel spreadsheet ARS-Media for Excel application. ARS-Media for Excel Instruction Manual is provided as a pdf file....

  20. GERMINAL — A computer code for predicting fuel pin behaviour

    Science.gov (United States)

    Melis, J. C.; Roche, L.; Piron, J. P.; Truffert, J.

    1992-06-01

    In the frame of the R and D on FBR fuels, CEA/DEC is developing the computer code GERMINAL to study the fuel pin thermal-mechanical behaviour during steady-state and incidental conditions. The development of GERMINAL is foreseen in two steps: (1) The GERMINAL 1 code designed as a "working horse" for immediate applications. The version 1 of GERMINAL 1 is presently delivered fully documented with a physical qualification guaranteed up to 8 at%. (2) The version 2 of GERMINAL 1, in addition to what is presently treated in GERMINAL 1 includes the treatment of high burnup effects on the fission gas release and the fuel-clad joint. This version, GERMINAL 1.2, is presently under testing and will be completed up to the end of 1991. The GERMINAL 2 code designed as a reference code for future applications will cover all the aspects of GERMINAL 1 (including high burnup effects) with a more general mechanical treatment, and a completely revised and advanced informatical structure.

  1. Hydrides reorientation investigation of high burn-up PWR fuel cladding

    Science.gov (United States)

    Valance, Stéphane; Bertsch, Johannes

    2015-09-01

    The direction of formation of hydride in fuel cladding tube is a major issue for the assessment of the cladding remaining ductility after service. This behavior is quite well known for fresh material, but few results exist for irradiated material. The reorientation behavior of a Zircaloy-4 fuel cladding (AREVA duplex DX-D4) at a burn-up of around 72 GWd t-1 is investigated here. The increase of the fraction of reoriented hydrides through repeated thermo-mechanical loading is inspected; as well, the possibility to recover a state with a minimized quantity of reoriented hydrides is tested using pure thermal loading cycles. The study is completed by a qualitative assessment of the hydrogen density in the duplex layer, where a dependence of the hydrides density on the hoop stress state is observed.

  2. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  3. Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  4. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Application

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Y. H.; Park, S. Y.; Lee, M. H.; Choi, B. K.; Baek, J. H.; Park, J. Y.; Kim, J. H.; Kim, H. G.; Jung, Y. H.; Bang, B. G

    2006-08-15

    The systematic study was performed to develop the advanced corrosion-resistant Zr alloys for high burnup and Gen IV application. The corrosion behavior was significantly changed with the alloy composition and the corrosion environment. In general, the model alloys with a higher alloying elements showed a higher corrosion resistance. Among the model alloys tested in this study, Zr-10Cr-0.2Fe showed the best corrosion resistance regardless of the corrosion condition. The oxide on the higher corrosion-resistant alloy such as Zr-1.0Cr-0.2Fe consisted of mainly columnar grains, and it have a higher tetragonal phase stability. In comparison with other alloys being considered for the SCWR, the Zr alloys showed a lower corrosion rate than ferritic-martensitic steels. The results of this study imply that, at least from a corrosion standpoint, Zr alloys deserve consideration as potential cladding or structural materials in supercritical water cooled reactors.

  5. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-19

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 and Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.

  6. The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Hao, E-mail: jiangh@ornl.gov; Wang, Jy-An John; Wang, Hong

    2016-12-01

    Highlights: • To investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on its dynamic performance. • Flexural rigidity, EI = M/κ, estimated from FEA results were benchmarked with SNF dynamic experimental results, and used to evaluate interface bonding efficiency. • Interface bonding efficiency can significantly dictate the SNF system rigidity and the associated dynamic performance. • With consideration of interface bonding efficiency and fuel cracking, HBU SNF fuel property was estimated with SNF static and dynamic experimental data. - Abstract: Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets to the clad, which results in a reduction in composite rod system flexural rigidity. Therefore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.

  7. Synthesis of VERCORS and Phebus data in severe accident codes and applications.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.

    2010-04-01

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, testing and incorporation of improved radionuclide release models into the MELCOR severe accident code.

  8. 1st order mass balance model - Excel and GAMS

    OpenAIRE

    ALS-NSCORT,

    2004-01-01

    Provider Notes:This zipfile contains the Excel files and GAMS code for a solvable version of the NSCORT mass balance.unzip this in an ECN working directory 1. in model_june04.xls, read the intro sheet and update the working directory cell. 2. run the macro create

  9. Cyclone Codes

    OpenAIRE

    Schindelhauer, Christian; Jakoby, Andreas; Köhler, Sven

    2016-01-01

    We introduce Cyclone codes which are rateless erasure resilient codes. They combine Pair codes with Luby Transform (LT) codes by computing a code symbol from a random set of data symbols using bitwise XOR and cyclic shift operations. The number of data symbols is chosen according to the Robust Soliton distribution. XOR and cyclic shift operations establish a unitary commutative ring if data symbols have a length of $p-1$ bits, for some prime number $p$. We consider the graph given by code sym...

  10. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  11. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Sammenfatning af de mest væsentlige pointer fra hovedrapporten: Dokumentation og evaluering af Coding Class......Sammenfatning af de mest væsentlige pointer fra hovedrapporten: Dokumentation og evaluering af Coding Class...

  12. Excel spreadsheet in teaching numerical methods

    Science.gov (United States)

    Djamila, Harimi

    2017-09-01

    One of the important objectives in teaching numerical methods for undergraduates’ students is to bring into the comprehension of numerical methods algorithms. Although, manual calculation is important in understanding the procedure, it is time consuming and prone to error. This is specifically the case when considering the iteration procedure used in many numerical methods. Currently, many commercial programs are useful in teaching numerical methods such as Matlab, Maple, and Mathematica. These are usually not user-friendly by the uninitiated. Excel spreadsheet offers an initial level of programming, which it can be used either in or off campus. The students will not be distracted with writing codes. It must be emphasized that general commercial software is required to be introduced later to more elaborated questions. This article aims to report on a teaching numerical methods strategy for undergraduates engineering programs. It is directed to students, lecturers and researchers in engineering field.

  13. code {poems}

    Directory of Open Access Journals (Sweden)

    Ishac Bertran

    2012-08-01

    Full Text Available "Exploring the potential of code to communicate at the level of poetry," the code­ {poems} project solicited submissions from code­writers in response to the notion of a poem, written in a software language which is semantically valid. These selections reveal the inner workings, constitutive elements, and styles of both a particular software and its authors.

  14. Designing Transformations: Schools of excellence

    Science.gov (United States)

    Marie Fairburn, Susan

    2011-12-01

    For over 50 years, including some form of Space education in school curriculum has become an established approach for inspiring young minds to study the sciences and pursue science-based careers. Space camps and schools are active all around the globe and typically attract the 'best and the brightest' young minds. But the context of Space is broad and all young minds need the confidence and ability to make choices that will best serve them in the world of learning, life and work. Designing Transformations: can the context of 'Space' serve as an inclusive educational engagement model and career skills tool to achieve "Schools of Excellence" (SoE)? This paper presents a case study for democratising space education within Scotland; through a new Schools of Excellence Model, which uses design methodology to convey an educational experience. It further demonstrates how the SoE model proposes the context of Space to dovetail with Scotland's Curriculum for Excellence (CfE) , with particular relevance to skills for learning, life and work to build skills and foster career agility to equip individuals for the new and changing demands of the future workplace. Curriculum for Excellence (CfE) is a new approach to the curriculum, with its emphasis on outcomes rather than inputs.

  15. Leadership, excellence, creativity and innovation

    OpenAIRE

    Coulson-Thomas, Colin

    2016-01-01

    Raises questions about the meaning, purpose and practice of contemporary leadership in relation to excellence, creativity and innovation, covering leadership qualities, the context and requirements of leadership, leadership at different stages of development, creativity and innovation, CEOs and top down leadership, entrepreneurship and shared leadership, leading the network organisation, shared and collective leadership, the role and contribution of boards, key questions for boards, leadershi...

  16. Patient satisfaction: focusing on "excellent".

    Science.gov (United States)

    Otani, Koichiro; Waterman, Brian; Faulkner, Kelly M; Boslaugh, Sarah; Burroughs, Thomas E; Dunagan, W Claiborne

    2009-01-01

    In an emerging competitive market such as healthcare, managers should focus on achieving excellent ratings to distinguish their organization from others. When it comes to customer loyalty, "excellent" has a different meaning. Customers who are merely satisfied often do not come back. The purpose of this study was to find out what influences adult patients to rate their overall experience as "excellent." The study used patient satisfaction data collected from one major academic hospital and four community hospitals. After conducting a multiple logistic regression analysis, certain attributes were shown to be more likely than others to influence patients to rate their experiences as excellent. The study revealed that staff care is the most influential attribute, followed by nursing care. These two attributes are distinctively stronger drivers of overall satisfaction than are the other attributes studied (i.e., physician care, admission process, room, and food). Staff care and nursing care are under the control of healthcare managers. If improvements are needed, they can be accomplished through training programs such as total quality management or continuous quality improvement, through which staff employees and nurses learn to be sensitive to patients' needs. Satisfying patients' needs is the first step toward having loyal patients, so hospitals that strive to ensure their patients are completely satisfied are more likely to prosper.

  17. Customer satisfaction and business excellence

    DEFF Research Database (Denmark)

    Kristensen, Kai; Martensen, Anne; Grønholdt, Lars

    The topic for this paper is the link between customer satisfaction and business performance, which makes it possible to use customer satisfaction measures as basis for creating business excellence. First, the paper presents microeconomic models for the relationship between customer satisfaction...

  18. Quality Management and Business Excellence

    Directory of Open Access Journals (Sweden)

    Vasile Dinu

    2017-02-01

    Full Text Available An excellent organization involves much more than the implementation and the certification of one or more models of management systems. It means developing techniques and tools of busin excellence which lead the organization to outstanding performance on quality, costs and deadlines in order to meet the expectations of all their stakeholders. Such an approach is needed especially in the context of an economy marked by globalization, extremely complex and dynamic that causes spectacular changes in the business environment by integrating quality management principles on purpose to develop sustainable excellence. Not coincidentally, the new edition of the European excellence model EFQM integrates for the first time the principle "managing with agility“ with the principles: “developing organizational capability”, “harnessing creativity & innovation”, “adding value to the customer”, “sustaining outstanding results” for the organization and “creating a sustainable future”. Also, the new model for quality management system defined by the edition from 2015 of ISO 9000 standards , promotes the process-based approach, incorporating the cycle "Plan - Do − Check − Act" (PDCA and the risk-based thinking, focusing on organizational change and innovation, in order to ensure a sustainable performance in business. Noteworthy is the endeavor for the development of a high-level structure for all international standards for management systems, aiming to harmonize these standards to facilitate the implementation of integrated management systems (quality − environment − security − social responsibility.

  19. 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)

  20. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  1. Coolant Density and Control Blade History Effects in Extended BWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Marshall, William BJ J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Martinez-Gonzalez, Jesus S [ORNL

    2015-01-01

    Oak Ridge National Laboratory and the US Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and predicted spent fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date investigating some aspects of extended BUC. (The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper.) Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC: (1) the effect of axial void profile and (2) the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of a modern operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. Although a single cycle does not provide complete data, the data obtained are sufficient to determine the primary effects and to identify conservative modeling approaches. These data were used in a study of the effect of axial void profile. The first stage of the study was determination of the necessary moderator density temporal fidelity in depletion modeling. After the required temporal fidelity was established, multiple void profiles were used to examine the effect on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied was control blade history. Control blades are inserted in

  2. Extended Burnup Credit for BWR Spent Nuclear Fuel in Storage and Transportation Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Martinez, J. S. [Univ. Politecnica de Madrid (Spain). Dept. of Nuclear Engineering

    2015-01-01

    [Full Text] Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate the application of burnup credit (BUC) for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase investigates the applicability of peak reactivity methods currently used for spent fuel pools to spent fuel storage and transportation casks and the validation of reactivity (keff) calculations and depleted fuel compositions. The second phase focuses on extending BUC beyond peak reactivity. This paper documents work performed to date, investigating some aspects of extended BUC, and it also describes the plan to complete the evaluations. The technical basis for application of peak reactivity methods to BWR fuel in storage and transportation systems is presented in a companion paper. Two reactor operating parameters are being evaluated to establish an adequate basis for extended BWR BUC, including investigation of the axial void profile effect and the effect of control blade utilization during operation. A detailed analysis of core simulator data for one cycle of an operating BWR plant was performed to determine the range of void profiles and the variability of the profile experienced during irradiation. While a single cycle does not provide complete data, the data obtained are sufficient to use to determine the primary effects and identify conservative modeling approaches. Using data resulting from a single cycle, the axial void profile is studied by first determining the temporal fidelity necessary in depletion modeling, and then using multiple void profiles to examine the effect of the void profile on cask reactivity. The results of these studies are being used to develop recommendations for conservatively modeling the void profile effects for BWR depletion calculations. The second operational parameter studied is control blade exposure. Control blades

  3. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  4. FRAPCON-1: a computer code for the steady state analysis of oxide fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Berna, G. A.; Bohn, M. P.; Coleman, D. R.; Lanning, D. D.

    1978-08-01

    FRAPCON is a FORTRAN IV computer code which predicts the steady state long-term burnup response of a light water reactor fuel rod. The coupled effects of fuel and cladding deformation, temperature, and internal gas pressure on the behavior of the fuel rod are considered in determining fuel rod response. The cladding deformation model includes multi-axial, elasto-plastic analysis and considers both primary and secondary creep. The fuel temperature model considers the effects of fuel cracking and relocation in determining the fuel temperature distribution. Burnup dependent fission gas generation and release is included in calculating fuel rod internal pressure. An integral fuel rod failure subcode determines failure and failure modes based on the operating conditions at each timestep. The material property subcode, MATPRO, provides gas, fuel and cladding properties to the computational subcodes in FRAPCON. No material properties need to be supplied by the code user. FRAPCON is a completely modular code with each major computational subcode isolated within the code and coupled to the main code by subroutine calls and data transfer through argument lists. FRAPCON is soft-coupled to the transient fuel rod code, FRAP-T, to provide initial conditions to initiate analysis of such off-normal transients as a loss-of-coolant accident. The code is presently programmed and running on a CDC 7600 computer.

  5. Sharing code.

    Science.gov (United States)

    Kubilius, Jonas

    2014-01-01

    Sharing code is becoming increasingly important in the wake of Open Science. In this review I describe and compare two popular code-sharing utilities, GitHub and Open Science Framework (OSF). GitHub is a mature, industry-standard tool but lacks focus towards researchers. In comparison, OSF offers a one-stop solution for researchers but a lot of functionality is still under development. I conclude by listing alternative lesser-known tools for code and materials sharing.

  6. Analog Coding.

    Science.gov (United States)

    CODING, ANALOG SYSTEMS), INFORMATION THEORY, DATA TRANSMISSION SYSTEMS , TRANSMITTER RECEIVERS, WHITE NOISE, PROBABILITY, ERRORS, PROBABILITY DENSITY FUNCTIONS, DIFFERENTIAL EQUATIONS, SET THEORY, COMPUTER PROGRAMS

  7. Effects of microstructural constraints on the transport of fission products in uranium dioxide at low burnups

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert [Arizona State University, Tempe, AZ (United States); Dickerson, Patricia [Los Alamos National Lab, Los Alamos, NM (United States); Gong, Bowen [Arizona State University, Tempe, AZ (United States); Peralta, Pedro, E-mail: pperalta@asu.edu [Arizona State University, Tempe, AZ (United States)

    2016-08-15

    Diffusion of fission gases in UO{sub 2} is studied at low burnups, before bubble growth and coalescence along grain boundaries (GBs) become dominant, using a 3-D finite element model that incorporates actual UO{sub 2} microstructures. Grain boundary diffusivities are assigned based on crystallography with lattice and GB diffusion coupled with temperature to account for temperature gradients. Heterogeneity of GB properties and connectivity can induce regions where concentration is locally higher than without GB diffusion. These regions are produced by “bottlenecks” in the GB network because of lack of connectivity among high diffusivity GBs due to crystallographic constraints, and they can lead to localized swelling. Effective diffusivities were calculated assuming a uniform distribution of high diffusivity among GBs. Results indicate an increase over the bulk diffusivity with a clear grain size effect and that connectivity and properties of different GBs become important factors on the variability of fission product concentration at the microscale. - Highlights: • Microstructure models are developed to study fission gas transport in oxide fuels. • Crystallographic and temperature dependent material properties are applied. • Fission product concentration is affected by grain boundary distribution. • High concentration regions can form as controlled by the grain boundary connectivity.

  8. Burnup simulations and spent fuel characteristics of ZrO 2 based inert matrix fuels

    Science.gov (United States)

    Schneider, E. A.; Deinert, M. R.; Herring, S. T.; Cady, K. B.

    2007-03-01

    Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF's) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO 2 based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.

  9. TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

    2014-04-01

    As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

  10. Chemical burnup determination based on spectrophotometric measurement of total rare earth fission products, uranium, and plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Marsh, S.F.; Ortiz, M.R.; Rein, J.E.

    1975-10-01

    A chemical burnup procedure incorporates the ion-exchange separation of uranium, plutonium, and total rare earth fission products (as the fission monitor) followed by the spectrophotometric determination of each. The separation involves retaining uranyl and plutonyl chloride complexes on a macroporous anion exchange column from 12 M HCl, whereas the rare earths and most fission products pass through. Subsequently, plutonium is eluted with 0.1 M HI-12 M HCl and uranium with 0.1 M HCl. From the initial effluent of the first column, the rare earth group is separated on a second column of either (1) macroporous anion exchange resin from HNO/sub 3/-CH/sub 3/OH, or (2) pellicular cation exchange particles from HCl-C/sub 2/H/sub 5/OH. The HNO/sub 3/--CH/sub 3/OH system normally is used to separate the rare earth group from fuel cladding elements and other fission products. The HCl--C/sub 2/H/sub 5/OH system additionally separates the rare earth group from americium. Arsenazo III is the chromogenic agent for the spectrophotometric determination of the separated uranium, plutonium, and rare earth fractions.

  11. SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

    Energy Technology Data Exchange (ETDEWEB)

    Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

    2014-04-01

    During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

  12. Thermal property change of MOX and UO{sub 2} irradiated up to high burnup of 74 GWd/t

    Energy Technology Data Exchange (ETDEWEB)

    Nakae, Nobuo, E-mail: nakae-nobuo@jnes.go.jp [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization (JNES), Toranomon Towers Office, 4-1-28, Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Kurematsu, Shigeru; Kosaka, Yuji [Nuclear Development Corporation (NDC), 622-12, Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Yoshino, Aya; Kitagawa, Takaaki [Mitsubishi Nuclear Fuel Co., LTD. (MNF), 12-1, Yurakucho 1-Chome, Chiyoda-ku, Tokyo 100-0006 (Japan)

    2013-09-15

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO{sub 2} fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO{sub 2}. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO{sub 2} is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO{sub 2} at high burnup under the condition that the pellet–cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO{sub 2} before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO{sub 2}. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  13. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  14. The Quest for Educational Excellence

    Directory of Open Access Journals (Sweden)

    Kathleen A. Tobin

    2015-04-01

    Full Text Available A review of the book “The Global Fourth Way: The Quest for Educational Excellence.” By Andy Hargreaves and Dennis Shirley. Thousand Oaks, CA: Corwin, 2012. ISBN: 978-1412987868 Individual chapters in The Global Fourth Way center on examples of high performance in Finland, Singapore, Alberta (Canada, Ontario (Canada, England, and California (US. Each captures the essence of educational models, practices, and issues in the region at large, before focusing on specific cases that support key values in their proposal.

  15. Commitment to project management excellence

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1994-12-31

    The paper discusses the commitment of the United States Department of Energy (DOE) to excellence in developing and implementing new project management and project control guidelines and in developing a training course. For the new Project Control Systems (PCS) Guidelines, DOE has consistently applied the precepts of Total Quality Management in the way problems were identified and resolved, in the use of techniques to develop the new guidance, and in the use of methods to create the new manager training. With the guidelines, DOE has responded to administration policy in developing a cost-effective, streamlined, and quantitative performance measurement. 3 refs.

  16. Excel-Based Business Analysis

    CERN Document Server

    Anari, Ali

    2012-01-01

    ai"The trend is your friend"is a practical principle often used by business managers, who seek to forecast future sales, expenditures, and profitability in order to make production and other operational decisions. The problem is how best to identify and discover business trends and utilize trend information for attaining objectives of firms.This book contains an Excel-based solution to this problem, applying principles of the authors' "profit system model" of the firm that enables forecasts of trends in sales, expenditures, profits and other business variables. The program,

  17. Divergence coding for convolutional codes

    Directory of Open Access Journals (Sweden)

    Valery Zolotarev

    2017-01-01

    Full Text Available In the paper we propose a new coding/decoding on the divergence principle. A new divergent multithreshold decoder (MTD for convolutional self-orthogonal codes contains two threshold elements. The second threshold element decodes the code with the code distance one greater than for the first threshold element. Errorcorrecting possibility of the new MTD modification have been higher than traditional MTD. Simulation results show that the performance of the divergent schemes allow to approach area of its effective work to channel capacity approximately on 0,5 dB. Note that we include the enough effective Viterbi decoder instead of the first threshold element, the divergence principle can reach more. Index Terms — error-correcting coding, convolutional code, decoder, multithreshold decoder, Viterbi algorithm.

  18. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...... development, Speaking Code unfolds an argument to undermine the distinctions between criticism and practice, and to emphasize the aesthetic and political aspects of software studies. Not reducible to its functional aspects, program code mirrors the instability inherent in the relationship of speech......; alternatives to mainstream development, from performances of the live-coding scene to the organizational forms of commons-based peer production; the democratic promise of social media and their paradoxical role in suppressing political expression; and the market’s emptying out of possibilities for free...

  19. 75 FR 48662 - Equity and Excellence Commission

    Science.gov (United States)

    2010-08-11

    ... Equity and Excellence Commission AGENCY: Office for Civil Rights, U.S. Department of Education. ACTION: Request for Nominations to Serve on the Equity and Excellence Commission. SUMMARY: The Secretary of... Equity and Excellence Commission. SUPPLEMENTARY INFORMATION: The Equity and Excellence Commission (Equity...

  20. Annual report of nuclear code evaluation committee for fiscal 2000 year

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-03-01

    In this report, research results discussed in fiscal 2000 year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. In 2000, papers mainly on the three topics of (1) present status of burnup credit evaluation methods, (2) issues concerning convergence of criticality calculation and (3) estimation methods for errors associated with criticality calculation based on nuclear data covariance file, are presented and discussed. These results are sorted to grasp the present status of related technology and described in this report. (author)

  1. Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.

  2. (Center of excellence: Microlaser microscope)

    Energy Technology Data Exchange (ETDEWEB)

    Webb, R.H.

    1992-01-01

    This Center-of-Excellence grant has two components: development of an imaging system based on microlaser arrays forms a central project among a group of laser diagnostic and therapeutic efforts primarily funded outside the grant. In these first 8 months we have set up the Microlaser Microscope using small microlaser arrays. We have emphasized the basics of microlaser handling and electronic addressing and the optics of the microscope. Details of electronics and optics given here will be used in the larger arrays which should be available soon. After a description of the central Microlaser Microscope project, we touch briefly on the other projects of the Center, which have been outstandingly fruitful this year. Publications are necessarily concerned with the smaller projects, since the Microlaser Microscope is in its early stages.

  3. Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    William J. Carmack

    2012-05-01

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in

  4. Some uncertainty results obtained by the statistical version of the KARATE code system related to core design and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.

    2017-11-15

    The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.

  5. Multicode comparison of selected source-term computer codes

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Parks, C.V.; Renier, J.P.; Roddy, J.W.; Ashline, R.C.; Wilson, W.B.; LaBauve, R.J.

    1989-04-01

    This report summarizes the results of a study to assess the predictive capabilities of three radionuclide inventory/depletion computer codes, ORIGEN2, ORIGEN-S, and CINDER-2. The task was accomplished through a series of comparisons of their output for several light-water reactor (LWR) models (i.e., verification). Of the five cases chosen, two modeled typical boiling-water reactors (BWR) at burnups of 27.5 and 40 GWd/MTU and two represented typical pressurized-water reactors (PWR) at burnups of 33 and 50 GWd/MTU. In the fifth case, identical input data were used for each of the codes to examine the results of decay only and to show differences in nuclear decay constants and decay heat rates. Comparisons were made for several different characteristics (mass, radioactivity, and decay heat rate) for 52 radionuclides and for nine decay periods ranging from 30 d to 10,000 years. Only fission products and actinides were considered. The results are presented in comparative-ratio tables for each of the characteristics, decay periods, and cases. A brief summary description of each of the codes has been included. Of the more than 21,000 individual comparisons made for the three codes (taken two at a time), nearly half (45%) agreed to within 1%, and an additional 17% fell within the range of 1 to 5%. Approximately 8% of the comparison results disagreed by more than 30%. However, relatively good agreement was obtained for most of the radionuclides that are expected to contribute the greatest impact to waste disposal. Even though some defects have been noted, each of the codes in the comparison appears to produce respectable results. 12 figs., 12 tabs.

  6. FY14 Status Report: CIRFT Testing Results on High Burnup UNF

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Jiang, Hao [ORNL

    2014-09-01

    The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

  7. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  8. Coding labour

    National Research Council Canada - National Science Library

    McCosker, Anthony; Milne, Esther

    2014-01-01

    ... software. Code encompasses the laws that regulate human affairs and the operation of capital, behavioural mores and accepted ways of acting, but it also defines the building blocks of life as DNA...

  9. Light water reactor fuel analysis code FEMAXI-V (Ver.1)

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-09-01

    A light water fuel analysis code FEMAXI-V is an advanced version which has been produced by integrating FEMAXI-IV(Ver.2), high burn-up fuel code EXBURN-I, and a number of functional improvements and extensions, to predict fuel rod behavior in normal and transient (not accident) conditions. The present report describes in detail the basic theories and structure, models and numerical solutions applied, improvements and extensions, and the material properties adopted in FEMAXI-V(Ver.1). FEMAXI-V deals with a single fuel rod. It predicts thermal and mechanical response of fuel rod to irradiation, including FP gas release. The thermal analysis predicts rod temperature distribution on the basis of pellet heat generation, changes in pellet thermal conductivity and gap thermal conductance, (transient) change in surface heat transfer to coolant, using radial one-dimensional geometry. The heat generation density profile of pellet can be determined by adopting the calculated results of burning analysis code. The mechanical analysis performs elastic/plastic, creep and PCMI calculations by FEM. The FP gas release model calculates diffusion of FP gas atoms and accumulation in bubbles, release and increase in internal pressure of rod. In every analysis, it is possible to allow some materials properties and empirical equations to depend on the local burnup or heat flux, which enables particularly analysis of high burnup fuel behavior and boiling transient of BWR rod. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  10. Reactivity loss validation of high burn-up PWR fuels with pile-oscillation experiments in MINERVE

    Energy Technology Data Exchange (ETDEWEB)

    Leconte, P.; Vaglio-Gaudard, C.; Eschbach, R.; Di-Salvo, J.; Antony, M.; Pepino, A. [CEA, DEN, DER, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    The ALIX experimental program relies on the experimental validation of the spent fuel inventory, by chemical analysis of samples irradiated in a PWR between 5 and 7 cycles, and also on the experimental validation of the spent fuel reactivity loss with bum-up, obtained by pile-oscillation measurements in the MINERVE reactor. These latter experiments provide an overall validation of both the fuel inventory and of the nuclear data responsible for the reactivity loss. This program offers also unique experimental data for fuels with a burn-up reaching 85 GWd/t, as spent fuels in French PWRs never exceeds 70 GWd/t up to now. The analysis of these experiments is done in two steps with the APOLLO2/SHEM-MOC/CEA2005v4 package. In the first one, the fuel inventory of each sample is obtained by assembly calculations. The calculation route consists in the self-shielding of cross sections on the 281 energy group SHEM mesh, followed by the flux calculation by the Method Of Characteristics in a 2D-exact heterogeneous geometry of the assembly, and finally a depletion calculation by an iterative resolution of the Bateman equations. In the second step, the fuel inventory is used in the analysis of pile-oscillation experiments in which the reactivity of the ALIX spent fuel samples is compared to the reactivity of fresh fuel samples. The comparison between Experiment and Calculation shows satisfactory results with the JEFF3.1.1 library which predicts the reactivity loss within 2% for burn-up of {approx}75 GWd/t and within 4% for burn-up of {approx}85 GWd/t. (authors)

  11. ITMO Photonics: center of excellence

    Science.gov (United States)

    Voznesenskaya, Anna; Bougrov, Vladislav; Kozlov, Sergey; Vasilev, Vladimir

    2016-09-01

    ITMO University, the leading Russian center in photonics research and education, has the mission to train highlyqualified competitive professionals able to act in conditions of fast-changing world. This paradigm is implemented through creation of a strategic academic unit ITMO Photonics, the center of excellence concentrating organizational, scientific, educational, financial, laboratory and human resources. This Center has the following features: dissemination of breakthrough scientific results in photonics such as advanced photonic materials, ultrafast optical and quantum information, laser physics, engineering and technologies, into undergraduate and graduate educational programs through including special modules into the curricula and considerable student's research and internships; transformation of the educational process in accordance with the best international educational practices, presence in the global education market in the form of joint educational programs with leading universities, i.e. those being included in the network programs of international scientific cooperation, and international accreditation of educational programs; development of mechanisms for the commercialization of innovative products - results of scientific research; securing financial sustainability of research in the field of photonics of informationcommunication systems via funding increase and the diversification of funding sources. Along with focusing on the research promotion, the Center is involved in science popularization through such projects as career guidance for high school students; interaction between student's chapters of international optical societies; invited lectures of World-famous experts in photonics; short educational programs in optics, photonics and light engineering for international students; contests, Olympics and grants for talented young researchers; social events; interactive demonstrations.

  12. Effect of fuel burnup and cross sections on modular HTGR (High-Temperature Gas-cooled Reactor) reactivity coefficients

    Science.gov (United States)

    Lefler, W.; Baxter, A.; Mathews, D.

    1987-12-01

    The temperature dependence of the reactivity coefficient in a prismatic block Modular High-Temperature Gas-Cooled Reactor (MHTGR) design is examined and found to be large and negative. Temperature coefficient results obtained with the ENDF/B-V data library were almost the same as results obtained with the earlier versions of the ENDF/B data library usually used at GA Technologies Inc., in spite of a significant eigenvalue increase with the ENDF/B-V data. The effects of fuel burnup and arbitrarily assumed cross section variations were examined and tabulated.

  13. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  14. Study of the triton-burnup process in different JET scenarios using neutron monitor based on CVD diamond

    Energy Technology Data Exchange (ETDEWEB)

    Nemtsev, G., E-mail: g.nemtsev@iterrf.ru; Amosov, V.; Meshchaninov, S.; Rodionov, R. [Institution “Project center ITER,” Moscow (Russian Federation); Popovichev, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Collaboration: EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2016-11-15

    We present the results of analysis of triton burn-up process using the data from diamond detector. Neutron monitor based on CVD diamond was installed in JET torus hall close to the plasma center. We measure the part of 14 MeV neutrons in scenarios where plasma current varies in a range of 1-3 MA. In this experiment diamond neutron monitor was also able to detect strong gamma bursts produced by runaway electrons arising during the disruptions. We can conclude that CVD diamond detector will contribute to the study of fast particles confinement and help predict the disruption events in future tokamaks.

  15. ALEPH2 - A general purpose Monte Carlo depletion code

    Energy Technology Data Exchange (ETDEWEB)

    Stankovskiy, A.; Van Den Eynde, G.; Baeten, P. [SCK CEN, Boeretang 200, B-2400 Mol (Belgium); Trakas, C.; Demy, P. M.; Villatte, L. [AREVA NP, Tour AREVA, Pl. J. Millier, 92084 Paris La Defense (France)

    2012-07-01

    The Monte-Carlo burn-up code ALEPH is being developed at SCK-CEN since 2004. A previous version of the code implemented the coupling between the Monte Carlo transport (any version of MCNP or MCNPX) and the ' deterministic' depletion code ORIGEN-2.2 but had important deficiencies in nuclear data treatment and limitations inherent to ORIGEN-2.2. A new version of the code, ALEPH2, has several unique features making it outstanding among other depletion codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. The last generation general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII and JENDL-4) are fully implemented, including special purpose activation, spontaneous fission, fission product yield and radioactive decay data. The built-in depletion algorithm allows to eliminate the uncertainties associated with obtaining the time-dependent nuclide concentrations. A predictor-corrector mechanism, calculation of nuclear heating, calculation of decay heat, decay neutron sources are available as well. The validation of the code on the results of REBUS experimental program has been performed. The ALEPH2 has shown better agreement with measured data than other depletion codes. (authors)

  16. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  17. Realisatie van Excel Kwadraat in de praktijk

    NARCIS (Netherlands)

    Dijkstra, Elma

    2013-01-01

    Dijkstra, E. M. (2013, 2 July). Realisatie van Excel Kwadraat in de praktijk [Realization of Excellent Education in practice]. Presentation held at the OLK-meeting, ITS, Radboud University, Nijmegen, The Netherlands.

  18. Excel for chemists: a comprehensive guide

    National Research Council Canada - National Science Library

    Billo, E. Joseph

    2011-01-01

    ..., biological, and medicinal calculations. Including a CD-ROM for Windows, this new edition provides chemists and students with the a detailed guide to using the current versions of Excel (Excel 2007 and 2010...

  19. Instant release of fission products in leaching experiments with high burn-up nuclear fuels in the framework of the Euratom project FIRST- Nuclides

    Science.gov (United States)

    Lemmens, K.; González-Robles, E.; Kienzler, B.; Curti, E.; Serrano-Purroy, D.; Sureda, R.; Martínez-Torrents, A.; Roth, O.; Slonszki, E.; Mennecart, T.; Günther-Leopold, I.; Hózer, Z.

    2017-02-01

    The instant release of fission products from high burn-up UO2 fuels and one MOX fuel was investigated by means of leach tests. The samples covered PWR and BWR fuels at average rod burn-up in the range of 45-63 GWd/tHM and included clad fuel segments, fuel segments with opened cladding, fuel fragments and fuel powder. The tests were performed with sodium chloride - bicarbonate solutions under oxidizing conditions and, for one test, in reducing Ar/H2 atmosphere. The iodine and cesium release could be partially explained by the differences in sample preparation, leading to different sizes and properties of the exposed surface areas. Iodine and cesium releases tend to correlate with FGR and linear power rating, but the scatter of the data is significant. Although the gap between the fuel and the cladding was closed in some high burn-up samples, fissures still provide possible preferential transport pathways.

  20. 77 FR 2711 - Equity and Excellence Commission

    Science.gov (United States)

    2012-01-19

    ... Equity and Excellence Commission AGENCY: Office for Civil Rights, Education. ACTION: Notice of open... Equity and Excellence Commission. This notice amends the January 5, 2012 notice by providing notice of a... Official, Equity and Excellence Commission, U.S. Department of Education, 400 Maryland Avenue SW...

  1. 76 FR 6774 - Equity and Excellence Commission

    Science.gov (United States)

    2011-02-08

    ... Equity and Excellence Commission AGENCY: Office for Civil Rights, Department of Education. ACTION: Notice... meeting of the Equity and Excellence Commission (Commission). The notice also describes the functions of... Chen, Designated Federal Official, Equity and Excellence Commission, U.S. Department of Education, 400...

  2. 75 FR 48661 - Equity and Excellence Commission

    Science.gov (United States)

    2010-08-11

    ... Equity and Excellence Commission AGENCY: U.S. Department of Education, Office for Civil Rights. ACTION: Notice of Establishment of the Equity and Excellence Commission. SUMMARY: The U.S. Secretary of Education (Secretary) announces the establishment of the Equity and Excellence Commission (Equity Commission or...

  3. 77 FR 484 - Equity and Excellence Commission

    Science.gov (United States)

    2012-01-05

    ... Equity and Excellence Commission AGENCY: Office for Civil Rights, U.S. Department of Education. ACTION...-coming meeting of the Equity and Excellence Commission (Commission). The notice also describes the..., Equity and Excellence Commission, U.S. Department of Education, 400 Maryland Avenue SW, Washington, DC...

  4. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  5. Network Coding

    Indian Academy of Sciences (India)

    Network coding is a technique to increase the amount of information °ow in a network by mak- ing the key observation that information °ow is fundamentally different from commodity °ow. Whereas, under traditional methods of opera- tion of data networks, intermediate nodes are restricted to simply forwarding their incoming.

  6. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  7. Network Coding

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 15; Issue 7. Network Coding. K V Rashmi Nihar B Shah P Vijay Kumar. General Article Volume 15 Issue 7 July 2010 pp 604-621. Fulltext. Click here to view fulltext PDF. Permanent link: http://www.ias.ac.in/article/fulltext/reso/015/07/0604-0621. Keywords.

  8. Expander Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Expander Codes - The Sipser–Spielman Construction. Priti Shankar. General Article Volume 10 ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science Bangalore 560 012, India.

  9. An Excel macro for generating trilinear plots.

    Science.gov (United States)

    Shikaze, Steven G; Crowe, Allan S

    2007-01-01

    This computer note describes a method for creating trilinear plots in Microsoft Excel. Macros have been created in MS Excel's internal language: Visual Basic for Applications (VBA). A simple form has been set up to allow the user to input data from an Excel worksheet. The VBA macro is used to convert the triangular data (which consist of three columns of percentage data) into X-Y data. The macro then generates the axes, labels, and grid for the trilinear plot. The X-Y data are plotted as scatter data in Excel. By providing this macro in Excel, users can create trilinear plots in a quick, inexpensive manner.

  10. High-Burnup-Structure (HBS): Model Development in MARMOT for HBS Formation and Stability Under Radiation and High Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Biner, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A detailed phase field model for the formation of High Burnup Structure (HBS) was developed and implemented in MARMOT. The model treats the HBS formation as an irradiation-induced recrystallization. The model takes into consideration the stored energy associated with dislocations formed under irradiation. The accumulation of radiation damage, hence, increases the system free energy and triggers recrystallization. The increase in the free energy due to the formation of new grain boundaries is offset by the reduction in the free energy by creating dislocation-free grains at the expense of the deformed grains. The model was first used to study the growth of recrystallized flat and circular grains. The model reults were shown to agree well with theorrtical predictions. The case of HBS formation in UO2 was then investigated. It was found that a threshold dislocation density of (or equivalently a threshold burn-up of 33-40 GWd/t) is required for HBS formation at 1200K, which is in good agrrement with theory and experiments. In future studies, the presence of gas bubbles and their effect on the formation and evolution of HBS will be considered.

  11. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Science.gov (United States)

    Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

    2017-09-01

    Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  12. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Directory of Open Access Journals (Sweden)

    Ternovykh Mikhail

    2017-01-01

    Full Text Available Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  13. Tree Coding of Bilevel Images

    DEFF Research Database (Denmark)

    Martins, Bo; Forchhammer, Søren

    1998-01-01

    probabilities to an arithmetic coder. The conditional probabilities are estimated from co-occurrence statistics of past pixels, the statistics are stored in a tree. By organizing the code length calculations properly, a vast number of possible models (trees) reflecting different pixel orderings can...... be investigated within reasonable time prior to generating the code. A number of general-purpose coders are constructed according to this principle. Rissanen's (1989) one-pass algorithm, context, is presented in two modified versions. The baseline is proven to be a universal coder. The faster version, which...... is one order of magnitude slower than JBIG, obtains excellent and highly robust compression performance. A multipass free tree coding scheme produces superior compression results for all test images. A multipass free template coding scheme produces significantly better results than JBIG for difficult...

  14. Polar Codes

    Science.gov (United States)

    2014-12-01

    added by the decoder is K/ρ+Td. By the last assumption, Td and Te are both ≤ K/ρ, so the total latency added is between 2K/ρ and 4K /ρ. For example...better resolution near the decision point. Reference [12] showed that in decoding a (1024, 512) polar code, using 6-bit LLRs resulted in per- formance

  15. Preliminary comparison of transport codes applied to a second-generation PT-SCWR lattice

    Energy Technology Data Exchange (ETDEWEB)

    Farkas, R.; Nichita, E. [University of Ontario Institute of Technology, Oshawa, ON (Canada)

    2014-07-01

    The Pressure-Tube Supercritical Water Cooled Reactor (PT-SCWR) is being developed in Canada as Canada's contribution to the Generation IV International Forum (GIF). A two-dimensional lattice benchmark has been previously developed to assess the applicability of various lattice physics codes to the PT-SCWR design. This work summarizes the benchmark results for two lattice codes: Serpent, a stochastic transport code, and DRAGON, a deterministic transport code. Specifically, k-effective, spectrum and multi-group macroscopic cross-sections as a function of burnup are compared. Preliminary results show a 4 mk difference in the reactivity values and a maximum 6% difference in two-group macroscopic cross sections. (author)

  16. Diseño de carteras con Excel®

    Directory of Open Access Journals (Sweden)

    Piñol Espasa, José-Agustín

    2013-12-01

    Full Text Available An Excel® workbook, containing three worksheets, is presented. The file allows performing financial portfolios using incorporated functions and VBA code (Visual Basic for Applications. The user must select the basic positions he/she wants to combine, and the program will supply analytical and graphical information about the portfolio’s yield at maturity. The workbook, named Carteras_Opciones.xlsm, is useful for traders at option markets, for investors as simulation tool, for finance courses and for people who enjoy Excel® spreadsheets. It is possible to download the xlsm file from the author’s website at the University of Valencia: www.uv.es/pinyol.

  17. Development of a method for xenon determination in the microstructure of high burn-up nuclear fuel[Dissertation 17527

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, M. I

    2008-07-01

    In nuclear fuel, in approximately one quarter of the fissions, one of the two formed fission products is gaseous. These are mainly the noble gases xenon and krypton with isotopes of xenon contributing up to 90% of the product gases. These noble fission gases do not combine with other species, and have a low solubility in the normally used uranium oxide matrix. They can be dissolved in the fuel matrix or precipitate in nanometer-sized bubbles within the fuel grain, in micrometer-sized bubbles at the grain boundaries, and a fraction also precipitates in fuel pores, coming from fuel fabrication. A fraction of the gas can also be released into the plenum of the fuel rod. With increasing fission, and therefore burn-up, the ceramic fuel material experiences a transformation of its structure in the 'cooler' rim region of the fuel. A subdivision occurs of the original fuel grains of few microns size into thousands of small grains of sub-micron sizes. Additionally, larger pores are formed, which also leads into an increasing porosity in the fuel rim, called high burn-up structure. In this structure, only a small fraction of the fission gas remains in the matrix, the major quantity is said to accumulate in these pores. Because of this accumulation, the knowledge of the quantities of gas within these pores is of major interest in consideration to burn-up, fuel performance and especially for safety issues. In case of design based accidents, i.e. rapidly increasing temperature transients, the behavior of the fuel has to be estimated. Various analytical techniques have been used to determine the Xe concentration in nuclear fuel samples. The capabilities of EPMA (Electron Probe Micro-Analyser) and SIMS (Secondary Ion Mass Spectrometry) have been studied and provided some qualitative information, which has been used for determining Xe-matrix concentrations. First approaches combining these two techniques to estimate pore pressures have been recently reported. However

  18. Measuring research excellence in the EU

    DEFF Research Database (Denmark)

    Sørensen, Mads P.; Bloch, Carter Walter; Young, Mitchell

    In 2013, the European Union unveiled its new ‘Composite Indicator for Scientific and Technological Research Excellence’, marking a turning point in how excellence is understood and used in European policy. This is not an isolated occurrence; policy-based interest in excellence is growing all over...... the world. The heightened focus on excellence and in particular, attempts to define it through quantitative indicators can have important implications for research policy and for the conduct of research itself. This paper examines how the European Union’s understanding of excellence has evolved in recent...... years, from the presentation of the Lisbon strategy in 2000 to the current Europe 2020 strategy. We find a distinct shift in the understanding of excellence and how success in the knowledge-based economy should be achieved: in the early period, excellence is a fuzzy concept, intrinsically embedded...

  19. Convolutional-Code-Specific CRC Code Design

    OpenAIRE

    Lou, Chung-Yu; Daneshrad, Babak; Wesel, Richard D.

    2015-01-01

    Cyclic redundancy check (CRC) codes check if a codeword is correctly received. This paper presents an algorithm to design CRC codes that are optimized for the code-specific error behavior of a specified feedforward convolutional code. The algorithm utilizes two distinct approaches to computing undetected error probability of a CRC code used with a specific convolutional code. The first approach enumerates the error patterns of the convolutional code and tests if each of them is detectable. Th...

  20. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  1. What every engineer should know about excel

    CERN Document Server

    Holman, J P

    2006-01-01

    INTRODUCTIONGetting the Most from ExcelConventionsOutline of MISCELLANEOUS OPERATIONS IN EXCEL AND WORDIntroductionPrint Screen or Screen DumpCustom Keyboard Setup for Symbols in WordViewing or Printing Column and Row Headings and Gridlines in ExcelAssorted InstructionsMoving Objects in Small Increments (Nudging)Formatting Objects in Word, Including WrappingFormatting Objects in ExcelUse of Photo-Editing Software in Word, Including WrappingCopying Cell Formulas: Effect of Relative and Absolute AddressesCopying Formulas by Dragging the Fill HandleShortcut for Changing the Status of Cell Address

  2. Excel 2013 power programming with VBA

    CERN Document Server

    Walkenbach, John

    2013-01-01

    Maximize your Excel 2013 experience using VBA application development The new Excel 2013 boasts updated features, enhanced power, and new capabilities. Naturally, that means John Walkenbach returns with a new edition of his bestselling VBA Programming book and covers all the methods and tools you need to know in order to program with Excel. With this comprehensive guide, ""Mr. Spreadsheet"" shows you how to maximize your Excel experience using professional spreadsheet application development tips from his own personal bookshelf. Featuring a complete introduction to Visual Basic f

  3. Excel 2003 Power Programming with VBA

    CERN Document Server

    Walkenbach, John

    2008-01-01

    "Today, no accomplished Excel programmer can afford to be without John's book. The value of Excel 2003 Power Programming with VBA is double most other books-simultaneously the premier reference and best learning tool for Excel VBA."--Loren Abdulezer, Author of Excel Best Practices for BusinessEverything you need to know about:* Creating stellar UserForms and custom dialog box alternatives * Working with VBA subprocedures and function procedures * Incorporating event-handling and interactions with other applications * Building user-friendly toolbars, menus, and help systems * Manipulating files

  4. Technical Excellence: A Requirement for Good Engineering

    Science.gov (United States)

    Gill, Paul S.; Vaughan, William W.

    2008-01-01

    Technical excellence is a requirement for good engineering. Technical excellence has many different ways of expressing itself within engineering. NASA has initiatives that address the enhancement of the Agency's technical excellence and thrust to maintain the associated high level of performance by the Agency on current programs/projects and as it moves into the Constellation Program and the return to the Moon with plans to visit Mars. This paper addresses some of the key initiatives associated with NASA's technical excellence thrust. Examples are provided to illustrate some results being achieved and plans to enhance these initiatives.

  5. Domain Decomposition Strategy for Pin-wise Full-Core Monte Carlo Depletion Calculation with the Reactor Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Jingang Liang

    2016-06-01

    Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

  6. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  7. Improvement of JRR-4 core management code system

    Energy Technology Data Exchange (ETDEWEB)

    Izumo, H.; Watanabe, S.; Nagatomi, H.; Hori, N. [Department of Research Reactor, Tokai Research Establishment, Japan Atomic Energy Institute, Tokai, Ibaraki (Japan)

    2000-10-01

    In the modification of JRR-4, the fuel was changed from 93% high enrichment uranium aluminized fuel to 20% low enriched uranium silicide fuel in conformity with the framework of reduced enrichment program on JAERI research reactors. As changing of this, JRR-4 core management code system which estimates excess reactivity of core, fuel burn-up and so on, was improved too. It had been difficult for users to operate the former code system because its input-output form was text-form. But, in the new code system (COMMAS-JRR), users are able to operate the code system without using difficult text-form input. The estimation results of excess reactivity of JRR-4 LEU fuel core were showed very good agreements with the measured value. It is the strong points of this new code system to be operated simply by using the windows form pictures act on a personal workstation equip with the graphical-user-interface (GUI), and to estimate accurately the specific characteristics of the LEU core. (author)

  8. A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique and Three Types of Gamma-ray Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

    2011-05-01

    A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

  9. ExcelAutomat: a tool for systematic processing of files as applied to quantum chemical calculations

    Science.gov (United States)

    Laloo, Jalal Z. A.; Laloo, Nassirah; Rhyman, Lydia; Ramasami, Ponnadurai

    2017-07-01

    The processing of the input and output files of quantum chemical calculations often necessitates a spreadsheet as a key component of the workflow. Spreadsheet packages with a built-in programming language editor can automate the steps involved and thus provide a direct link between processing files and the spreadsheet. This helps to reduce user-interventions as well as the need to switch between different programs to carry out each step. The ExcelAutomat tool is the implementation of this method in Microsoft Excel (MS Excel) using the default Visual Basic for Application (VBA) programming language. The code in ExcelAutomat was adapted to work with the platform-independent open-source LibreOffice Calc, which also supports VBA. ExcelAutomat provides an interface through the spreadsheet to automate repetitive tasks such as merging input files, splitting, parsing and compiling data from output files, and generation of unique filenames. Selected extracted parameters can be retrieved as variables which can be included in custom codes for a tailored approach. ExcelAutomat works with Gaussian files and is adapted for use with other computational packages including the non-commercial GAMESS. ExcelAutomat is available as a downloadable MS Excel workbook or as a LibreOffice workbook.

  10. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  11. Concatenated codes with convolutional inner codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Thommesen, Christian; Zyablov, Viktor

    1988-01-01

    The minimum distance of concatenated codes with Reed-Solomon outer codes and convolutional inner codes is studied. For suitable combinations of parameters the minimum distance can be lower-bounded by the product of the minimum distances of the inner and outer codes. For a randomized ensemble...... of concatenated codes a lower bound of the Gilbert-Varshamov type is proved...

  12. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  13. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  14. Using Microsoft Excel to Generate Usage Statistics

    Science.gov (United States)

    Spellman, Rosemary

    2011-01-01

    At the Libraries Service Center, statistics are generated on a monthly, quarterly, and yearly basis by using four Microsoft Excel workbooks. These statistics provide information about what materials are being requested and by whom. They also give details about why certain requests may not have been filled. Utilizing Excel allows for a shallower…

  15. Getting the Best out of Excel

    Science.gov (United States)

    Heys, Chris

    2008-01-01

    Excel, Microsoft's spreadsheet program, offers several tools which have proven useful in solving some optimization problems that arise in operations research. We will look at two such tools, the Excel modules called Solver and Goal Seek--this after deriving an equation, called the "cash accumulation equation", to be used in conjunction with them.

  16. Production management using the EFQM Excellence Model

    Directory of Open Access Journals (Sweden)

    Janja Škedel

    2016-09-01

    Full Text Available Research Question (RQ: Comparison of production management using the EFQM Excellence Model. Purpose: The aim of the research is based on a comparison of production management using the EFQM Excellence Model, to establish identity and difference. The aim is to improve the management of production and by using the model closer to excellence. Method: An identification method of benchmarking. Results: The results show tolerance, which represent an opportunity to improve production management with excellence. Organization: If we take into account the results of the organization, this would be an asset to the organization. Society: Method comparisons can also be used in the wider environment. Originality: The survey is unique and the first of its kind in the manufacturing organization. Limitations/Future Research: With this research we will gain improvements in production management through design excellence.

  17. A Four Group Reference Code for Solving Neutron Diffusion Equation in a VVER-440 Core

    Energy Technology Data Exchange (ETDEWEB)

    Saarinen, Simo [Fortum Nuclear Services Ltd., P.O. Box 100, 00048 Fortum (Finland)

    2008-07-01

    Nuclear reactor core power calculation is essential in the analysis of the nuclear power plant and especially the core. Currently, the core power distribution in Loviisa VVER-440 core is calculated using nodal code HEXBU-3D and pin-power reconstruction code ELSI-1440 that solve the two group neutron diffusion equation. The computer power available has increased significantly during the last decades allowing us to develop a fine mesh code HEXRE for solving the four group diffusion equation. The diffusion equations are discretized using piecewise linear polynomials. The core is discretized using one node per fuel pin cell. The axial discretization can be chosen freely. The boundary conditions are described using diffusion theory and albedos. Burnup dependence is modelled by tabulating diffusion parameters at certain burnup values and using interpolation for the intermediate values. A two degree polynomial is used for the modelling of the feedback effects. Eigenvalue calculation for both boron concentration and multiplication factor control has been formulated. A possibility to perform fuel loading and shuffling operations is implemented. HEXRE has been thoroughly compared with HEXBU-3D and ELSI-1440. The effect of the different energy and space discretizations used is investigated. Some safety criteria for the core calculated with the HEXRE and HEXBU-3D/ELSI-1440 have been compared. From the calculations (e.g. the safety criteria) we can estimate whether there exists systematic deviations in HEXBU- 3D/ELSI-1440 calculations or not. (author)

  18. People management as indicator of business excellence

    DEFF Research Database (Denmark)

    Haffer, Rafal; Kristensen, Kai

    2010-01-01

    Purpose – This paper aims to show the importance of people management as a key indicator of business excellence based on four research projects, conducted on the samples of Polish (in the years 2004-2005 and 2006-2007) and Danish companies (in 1999 and 2005). Design/methodology/approach – EFQM...... it possible to compare developing Polish and developed Danish companies in their initiatives aiming at business excellence. Findings – The results indicate significant negligence in the management of human resources as one of the initiatives towards business excellence of Polish enterprises before Poland...

  19. Lineær programmering med Excel

    DEFF Research Database (Denmark)

    Lynggaard, Peter

    Publikationen giver en introduktion til faget "Lineær programmering". Formulering og løsning bygger på, at modellerne opstilles og løses i et Excel-regneark. Fremgangsmåden er forklaret trin for trin, således at hæftet kan bruges som selvstudiemateriale. Indholdsfortegnelse Generelt om styring og...... planlægning Lineær programmering Excel solveren Fortolkning af udskrifter Degeneration Excel tips Opgaver med løsninger...

  20. Statistical analysis with Excel for dummies

    CERN Document Server

    Schmuller, Joseph

    2013-01-01

    Take the mystery out of statistical terms and put Excel to work! If you need to create and interpret statistics in business or classroom settings, this easy-to-use guide is just what you need. It shows you how to use Excel's powerful tools for statistical analysis, even if you've never taken a course in statistics. Learn the meaning of terms like mean and median, margin of error, standard deviation, and permutations, and discover how to interpret the statistics of everyday life. You'll learn to use Excel formulas, charts, PivotTables, and other tools to make sense of everything fro

  1. 101 Ready-To-Use Excel Macros

    CERN Document Server

    Alexander, Michael

    2012-01-01

    Save time and be more productive with this helpful guide to Excel macros! While most books about Excel macros offer only minor examples, usually aimed at illustrating a particular topic, this invaluable resource provides you with the tools needed to efficiently and effectively program Excel macros immediately. Step-by-step instructions show you how to create VBA macros and explain how to customize your applications to look and work exactly as you want them to. By the end of the book, you will understand how each featured macro works, be able to reuse the macros included in the book and online,

  2. Microsoft Office Excel 2003 Inside Out

    CERN Document Server

    Stinson, Craig

    2003-01-01

    Hey, you know your way around a spreadsheet-so now dig into Excel 2003 and really put your data to work! This supremely organized reference packs hundreds of timesaving solutions, troubleshooting tips, and handy workarounds in concise, fast-answer format. It's all muscle and no fluff. Discover the best and fastest ways to perform everyday tasks, and challenge yourself to new levels of Excel mastery! Build on what you already know about Excel and quickly dive into what's newLearn how the experts design more powerful spreadsheetsSharpen your core to advanced document editing and formatting skil

  3. Advanced Excel for scientific data analysis

    CERN Document Server

    De Levie, Robert

    2004-01-01

    Excel is by far the most widely distributed data analysis software but few users are aware of its full powers. Advanced Excel For Scientific Data Analysis takes off from where most books dealing with scientific applications of Excel end. It focuses on three areas-least squares, Fourier transformation, and digital simulation-and illustrates these with extensive examples, often taken from the literature. It also includes and describes a number of sample macros and functions to facilitate common data analysis tasks. These macros and functions are provided in uncompiled, computer-readable, easily

  4. What Makes an Excellent Lecturer? Academics' Perspectives on the Discourse of "Teaching Excellence" in Higher Education

    Science.gov (United States)

    Wood, Margaret; Su, Feng

    2017-01-01

    In the context of the Teaching Excellence Framework (TEF), we examine academics' perspectives on the discourse of "teaching excellence" based on an empirical study with 16 participants from five post-1992 universities. The article reports the findings on academics' views of the term and concept of "teaching excellence",…

  5. Excellence in the Knowledge-Based Economy: From Scientific to Research Excellence

    Science.gov (United States)

    Sørensen, Mads P.; Bloch, Carter; Young, Mitchell

    2016-01-01

    In 2013, the European Union (EU) unveiled its new "Composite Indicator for Scientific and Technological Research Excellence." This is not an isolated occurrence; policy-based interest in excellence is growing all over the world. The heightened focus on excellence and, in particular, attempts to define it through quantitative indicators…

  6. Introduction to statistics using Microsoft Excel

    CERN Document Server

    Remenyi, Dan; English, Joseph

    2011-01-01

    This book explains the statistical concepts and then uses Microsoft Excel functions to illustrate how to get results using the appropriate techniques which will help researchers directly with their research.

  7. Predictors of excellent response to lithium

    DEFF Research Database (Denmark)

    Kessing, Lars Vedel; Hellmund, Gunnar; Andersen, Per Kragh

    2011-01-01

    The aim of this study was to identify sociodemographic and clinical predictors of excellent response, that is, 'cure' of future affective episodes, to lithium in monotherapy. We used nationwide registers to identify all patients with a diagnosis of bipolar disorder in psychiatric hospital settings...... who were prescribed lithium from 1995 to 2006 in Denmark (N=3762). Excellent lithium responders were defined as patients who after a stabilization lithium start-up period of 6 months, continued lithium in monotherapy without getting hospitalized. The rate of excellent response to lithium...... with somatic comorbidity had increased rates of non-response to lithium compared with patients without somatic comorbidity (HR=1.23, 95% CI: 1.00-1.52).It is concluded that the prevalence of excellent response to lithium monotherapy is low and such patients are characterized by few earlier psychiatric...

  8. Solving rational expectations models using Excel

    DEFF Research Database (Denmark)

    Strulik, Holger

    2004-01-01

    Problems of discrete time optimal control can be solved using backward iteration and Microsoft Excel. The author explains the method in general and shows how the basic models of neoclassical growth and real business cycles are solved...

  9. Faculty Evaluation Stimulates Expectations of Excellence.

    Science.gov (United States)

    Andrews, Hans A.; Marzano, William A.

    1984-01-01

    Considers important elements of comprehensive faculty evaluation systems (e.g., provision of formative feedback for instructors performing unsatisfactorily and of incentives for excellence). Details the development and results of the evaluation system at Illinois Valley Community College. (LAL)

  10. Editorial: South African Guidelines Excellence (SAGE): Efficient ...

    African Journals Online (AJOL)

    Editorial: South African Guidelines Excellence (SAGE): Efficient, effective and unbiased clinical practice guideline teams. Karen Grimmer, Janine Margarita Dizon, Quinette Louw, Tamara Kredo, Taryn Young, Shingai Machingaidze ...

  11. Contribution of Lean Management to Excellence

    Directory of Open Access Journals (Sweden)

    López-Fresno Palmira

    2014-11-01

    Full Text Available To continuously and systematically improve efficiency and efficacy of processes, organizations need the implication of all employees in continuous improvement and innovation through suitable Quality Management Programs (QMPs. Effectiveness of these programs is directly linked to the requirement employees understand the methodologies and tools used for QM and the benefits that will derivate from their implementation, individually and collectively, so they can commit and implicate. Lean Management is a friendly methodology to continuously and systematically achieve process improvement, so helping the organization seeking operational excellence that contributes to overall excellence. This paper identifies Critical Success Factors (CSFs for an effective implementation of QMPs, suggests Lean Management as an easy-to-understand, powerful and friendly methodology for operational excellence and overall excellence, and presents a case experience of implementation of Lean Management in a health care organization that applies the EFQM model, and the lessons learnt.

  12. The Business Excellence Model for CSR Implementation?

    DEFF Research Database (Denmark)

    Neergaard, Peter; Gjerdrum Pedersen, Esben Rahbek

    2012-01-01

    Most of the Fortune 500 companies address Corporate Social Responsibility (CSR) on their websites. However, CSR remains a fluffy concept difficult to implement in organization. The European Business Excellence Model has since the introduction in 1992 served as a powerful tool for integrating...... European Award winning company has used the model to integrate CSR. The company adapted the Business Excellence model to improve performance, stimulate innovation and consensus....

  13. Idea of Quality Versus Idea of Excellence

    Directory of Open Access Journals (Sweden)

    Marko Kiauta

    2012-12-01

    Full Text Available This study investigates professionals on the field of quality, are responsible to give to customer honest clarification of fundamental ideas. Quality movement is losing credibility with suggesting that the idea of quality is replacing with the idea of excellence. Findings are based on more than 25 years of practice in professional promotion of quality: in consulting on private and public sector, from 1990 lead auditor at SIQ (Slovenian Institute of Quality, from 1998 lead assessor – commission for Slovenian Excellence Quality Award. Theory is developed based on: Noriaki Kano theory of Attractive quality, Tito Conti ideas on TQM and applications problems of Excellence model, Practical case of General Hospital Novo Mesto (in 1998 first attempt of using EM, than forced to build QMS based on ISO 9001 and then returned to practice EM. Findings: We really need to amplify and to understand the concept of quality in a much wider way. To treat excellence related activities separated from all others quality management activities is not god solution. The name of EFQM Excellence Model should be replaced with Quality Management Model. Research limitations/implications: This paper present findings mainly based on practice in Slovenia and especially in public sector where practicing of CAF is not giving expected benefits. Practical implications: The three styles of quality management (improvements to reach demands, improvements to reach expectations, improvements to react on new conditions and needs should be connected with personal development. Theory is developed based on: Noriaki Kano theory of Attractive quality, Tito Conti ideas on TQM and applications problems of Excellence model. We need integration moments. Integration is other word for creativity and health. It leads to integrity. Excellence is only one of three states of quality. If we ask: How? The answer is bad, good or excellent. All three are possible states of the same parameter.

  14. On the use of SERPENT Monte Carlo code to generate few group diffusion constants

    Energy Technology Data Exchange (ETDEWEB)

    Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)

  15. Limitations of Using Microsoft Excel Version 2016 (MS Excel 2016) for Statistical Analysis for Medical Research.

    Science.gov (United States)

    Tanavalee, Chotetawan; Luksanapruksa, Panya; Singhatanadgige, Weerasak

    2016-06-01

    Microsoft Excel (MS Excel) is a commonly used program for data collection and statistical analysis in biomedical research. However, this program has many limitations, including fewer functions that can be used for analysis and a limited number of total cells compared with dedicated statistical programs. MS Excel cannot complete analyses with blank cells, and cells must be selected manually for analysis. In addition, it requires multiple steps of data transformation and formulas to plot survival analysis graphs, among others. The Megastat add-on program, which will be supported by MS Excel 2016 soon, would eliminate some limitations of using statistic formulas within MS Excel.

  16. 101 ready-to-use Excel formulas

    CERN Document Server

    Alexander, Michael

    2014-01-01

    Mr. Spreadsheet has done it again with 101 easy-to-apply Excel formulas 101 Ready-to-Use Excel Formulas is filled with the most commonly-used, real-world Excel formulas that can be repurposed and put into action, saving you time and increasing your productivity. Each segment of this book outlines a common business or analysis problem that needs to be solved and provides the actual Excel formulas to solve the problem-along with detailed explanation of how the formulas work. Written in a user-friendly style that relies on a tips and tricks approach, the book details how to perform everyday Excel tasks with confidence. 101 Ready-to-Use Excel Formulas is sure to become your well-thumbed reference to solve your workplace problems. The recipes in the book are structured to first present the problem, then provide the formula solution, and finally show how it works so that it can be customized to fit your needs. The companion website to the book allows readers to easily test the formulas and provides visual confirmat...

  17. Fundamentals of convolutional coding

    CERN Document Server

    Johannesson, Rolf

    2015-01-01

    Fundamentals of Convolutional Coding, Second Edition, regarded as a bible of convolutional coding brings you a clear and comprehensive discussion of the basic principles of this field * Two new chapters on low-density parity-check (LDPC) convolutional codes and iterative coding * Viterbi, BCJR, BEAST, list, and sequential decoding of convolutional codes * Distance properties of convolutional codes * Includes a downloadable solutions manual

  18. A computer code for calculation of radioactive nuclide generation and depletion, decay heat and {gamma} ray spectrum. FPGS90

    Energy Technology Data Exchange (ETDEWEB)

    Ihara, Hitoshi; Katakura, Jun-ichi; Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1995-11-01

    In a nuclear reactor radioactive nuclides are generated and depleted with burning up of nuclear fuel. The radioactive nuclides, emitting {gamma} ray and {beta} ray, play role of radioactive source of decay heat in a reactor and radiation exposure. In safety evaluation of nuclear reactor and nuclear fuel cycle, it is needed to estimate the number of nuclides generated in nuclear fuel under various burn-up condition of many kinds of nuclear fuel used in a nuclear reactor. FPGS90 is a code calculating the number of nuclides, decay heat and spectrum of emitted {gamma} ray from fission products produced in a nuclear fuel under the various kinds of burn-up condition. The nuclear data library used in FPGS90 code is the library `JNDC Nuclear Data Library of Fission Products - second version -`, which is compiled by working group of Japanese Nuclear Data Committee for evaluating decay heat in a reactor. The code has a function of processing a so-called evaluated nuclear data file such as ENDF/B, JENDL, ENSDF and so on. It also has a function of making figures of calculated results. Using FPGS90 code it is possible to do all works from making library, calculating nuclide generation and decay heat through making figures of the calculated results. (author).

  19. On the condition of UO2 nuclear fuel irradiated in a PWR to a burn-up in excess of 110 MWd/kgHM

    Science.gov (United States)

    Restani, R.; Horvath, M.; Goll, W.; Bertsch, J.; Gavillet, D.; Hermann, A.; Martin, M.; Walker, C. T.

    2016-12-01

    Post-irradiation examination results are presented for UO2 fuel from a PWR fuel rod that had been irradiated to an average burn-up of 105 MWd/kgHM and showed high fission gas release of 42%. The radial distribution of xenon and the partitioning of fission gas between bubbles and the fuel matrix was investigated using laser ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) and electron probe microanalysis. It is concluded that release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In this region thermal release had occurred from the high burn-up structure (HBS) at some point after the sixth irradiation cycle. The LA-ICP-MS results indicate that gas release had also occurred from the HBS in the vicinity of the pellet periphery. It is shown that the gas pressure in the HBS pores is well below the pressure that the fuel can sustain.

  20. Determination of uranium concentration and burn-up of irradiated reactor fuel in contaminated areas in Belarus using uranium isotopic ratios in soil samples

    OpenAIRE

    Mironov, V. P.; Matusevich, J. L.; Kudrjashov, V. P.; Ananich, P. I.; Zhuravkov, V. V.; Boulyga, S. F.; Becker, J. S.

    2005-01-01

    An analytical method is described for the estimation of uranium concentrations, of U-235/U-238 and U-236/U-238 isotope ratios and burn-up of irradiated reactor uranium in contaminated soil samples by inductively coupled plasma mass spectrometry. Experimental results obtained at 12 sampling sites situated on northern and western radioactive fallout tails 4 to 53 km distant from Chernobyl nuclear power plant (NPP) are presented. Concentrations of irradiated uranium in the upper 0-10cm soil laye...

  1. Rapid aqueous release of fission products from high burn-up LWR fuel: Experimental results and correlations with fission gas release

    Science.gov (United States)

    Johnson, L.; Günther-Leopold, I.; Kobler Waldis, J.; Linder, H. P.; Low, J.; Cui, D.; Ekeroth, E.; Spahiu, K.; Evins, L. Z.

    2012-01-01

    Studies of the rapid aqueous release of fission products from UO 2 and MOX fuel are of interest for the assessment of the safety of geological disposal of spent fuel, because of the associated potential contribution to dose in radiological safety assessment. Studies have shown that correlations between fission gas release (FGR) and the fraction rapidly leached of various long-lived fission products can provide a useful method to obtain some of this information. Previously, these studies have been limited largely to fuel with burn-up values below 50 MWd/kg U. Collaborative studies involving SKB, Studsvik, Nagra and PSI have provided new data on short-term release of 137Cs and 129I for a number of fuels irradiated to burn-ups of 50-75 MWd/kgU. In addition a method for analysis of leaching solutions for 79Se was developed. The results of the studies show that the fractional release of 137Cs is usually much lower than the FGR covering the entire range of burn-ups studied. Fractional 129I releases are somewhat larger, but only in cases in which the fuel was forcibly extracted from the cladding. Despite the expected high degree of segregation of fission gas (and by association 137Cs and 129I) in the high burn-up rim, no evidence was found for a significant contribution to release from the rim region. The method for 79Se analysis developed did not permit its detection. Nonetheless, based on the detection limit, the results suggest that 79Se is not preferentially leached from spent fuel.

  2. Creating cultures of excellence: Strategies and outcomes

    Directory of Open Access Journals (Sweden)

    Michael Mintrom

    2014-12-01

    Full Text Available Research findings on effective support for learning, the development of expertise, and the psychology of success suggest that the pursuit of excellence is teachable. Within the emerging field of research and practice termed “the scholarship of teaching and learning,” considerable effort has been made to document the practices of teachers who, by various measures, have been deemed excellent. In contrast, no effort has been made to codify how students can be trained to self-consciously build behaviors that generate excellent outcomes. This article reports on a multi-year effort to create cultures of excellence among cohorts of graduate students. A statistical analysis of subsequent student performance on a significant, related task indicates that explicitly promoting a culture of excellence among course participants can have a positive and sustained impact on their individual practices. Comments from subsequent student reflections further support this claim. The teaching strategies reported here could be refined, replicated, and reinvented to good effect across higher education. They are also of special relevance to those delivering professional development training to early- and mid-career professionals.

  3. AUS98 - The 1998 version of the AUS modular neutronic code system

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, G.S.; Harrington, B.V

    1998-07-01

    AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous AUS publications are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM main-frame computers to UNIX workstations This report gives details of all system aspects of AUS and all modules except the POW3D multi-dimensional diffusion module refs., tabs.

  4. Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

    1997-03-01

    The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

  5. Evaluation of the elastic constants of the high burnup nuclear fuel (U{sub 1-y},Gd{sub y})O{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Leite, Luciana T.; Dias, Marcio S., E-mail: ltl@cdtn.b, E-mail: marciod@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    Nuclear fuels with burnable poisons are applications in the scope of the nuclear energy for electricity generation. The discharge burnups of this type of fuel are higher than the discharge burnups of pure UO{sub 2} fuels with same enrichment of the fissile element. In conditions of high burnups and moderate transients, the fuel rods can be subject to the pellet/cladding mechanical interactions (PCMI). Elastic constants of the fuel and cladding are used in the thermo-mechanical evaluation of the PCMI. The fuel elastic constants are usually expressed as a function of temperature, composition and porosity. Data survey and modeling of elastic constants for the (U{sub 1-y},Gd{sub y})O{sub 2} fuel are developed in this paper. Due to small amount of the available measurements for the (U{sub 1-y},Gd{sub y})O{sub 2}, the modeling taken into account the data from UO{sub 2} and Gd{sub 2}O{sub 3} measurements. Available measurements of the Al{sub 2}O{sub 3} were also used in the model validation. The elastic constants of these materials have been fitted by an analytic equation, which depicts the whole range the dependences. (author)

  6. Blue code: Is it a real emergency?

    Science.gov (United States)

    Eroglu, Serkan E; Onur, Ozge; Urgan, Oğuz; Denizbasi, Arzu; Akoglu, Haldun

    2014-01-01

    Cardiac arrests in hospital areas are common, and hospitals have rapid response teams or "blue code teams" to reduce preventable in-hospital deaths. Education about the rapid response team has been provided in all hospitals in Turkey, but true "blue code" activation is rare, and it is abused by medical personnel in practice. This study aimed to determine the cases of wrong blue codes and reasons of misuse. This retrospective study analyzed the blue code reports issued by our hospital between January 1 and June 1 2012. A total of 89 "blue code" activations were recorded in 5 months. A "blue code" was defined as any patient with an unexpected cardiac or respiratory arrest requiring resuscitation and activation of a hospital alert. Adherence to this definition, each physician classified their collected activation forms as either a true or a wrong code. Then, patient data entered a database (Microsoft Excel 2007 software) which was pooled for analysis. The data were analyzed by using frequencies and the Chi-square test on SPSSv16.0. The patients were diagnosed with cardiopulmonary arrest (8), change in mental status (18), presyncope (11), chest pain (12), conversive disorder (18), and worry of the staff for the patient (22). Code activation was done by physicians in 76% of the patients; the most common reason for blue code was concern of staff for the patient. The findings of this study show that more research is needed to establish the overall effectiveness and optimal implementation of blue code teams.

  7. Switching Codes in the Plurilingual Classroom

    Science.gov (United States)

    Corcoll López, Cristina; González-Davies, Maria

    2016-01-01

    The English as a foreign language classroom is a plurilingual setting par excellence since it involves at least two languages. However, plurilingual practices such as code-switching and translation have been consistently discouraged in formal learning contexts, based on the belief that keeping languages compartmentalized helps learning, and…

  8. Excellence in the knowledge-based economy: from scientific to research excellence

    DEFF Research Database (Denmark)

    Sørensen, Mads P.; Bloch, Carter Walter; Young, Mitchell

    2016-01-01

    to define it through quantitative indicators can have important implications for research policy and for the conduct of research itself. This paper examines how the EU's understanding of excellence has evolved in recent years, from the presentation of the Lisbon strategy in 2000 to the current Europe 2020......In 2013, the European Union (EU) unveiled its new ‘Composite Indicator for Scientific and Technological Research Excellence’. This is not an isolated occurrence; policy-based interest in excellence is growing all over the world. The heightened focus on excellence and, in particular, attempts...... strategy. We find a distinct shift in the understanding of excellence and how success in the knowledge-based economy should be achieved: in the early period, excellence is a fuzzy concept, intrinsically embedded in research and researchers and revealed by peer review. In the later period, excellence...

  9. Auto Draw from Excel Input Files

    Science.gov (United States)

    Strauss, Karl F.; Goullioud, Renaud; Cox, Brian; Grimes, James M.

    2011-01-01

    The design process often involves the use of Excel files during project development. To facilitate communications of the information in the Excel files, drawings are often generated. During the design process, the Excel files are updated often to reflect new input. The problem is that the drawings often lag the updates, often leading to confusion of the current state of the design. The use of this program allows visualization of complex data in a format that is more easily understandable than pages of numbers. Because the graphical output can be updated automatically, the manual labor of diagram drawing can be eliminated. The more frequent update of system diagrams can reduce confusion and reduce errors and is likely to uncover symmetric problems earlier in the design cycle, thus reducing rework and redesign.

  10. Does excellence matter? The influence of potential for excellence on students’ motivation for specific collaborative tasks

    OpenAIRE

    Pullen, Annedien G.; Griffioen, Didi M. E.; Schoonenboom, Judith; Koning, Björn; Beishuizen, Jos J.

    2017-01-01

    textabstractIt is often assumed that students with a higher potential for excellence are less motivated to collaborate. So far, the question remains whether this is actually the case. This survey study investigated the influence of business students’ (N = 389) potential for excellence on their motivation to collaborate on a business-related task. Different aspects of potential for excellence were taken into account, including intelligence, creativity, first-year grade point average (GPA), and...

  11. Excel formulas and functions for dummies

    CERN Document Server

    Bluttman, Ken

    2013-01-01

    Learn to use Excel for practical, day-to-day calculations Excel is a powerful program with more than 300 built-in functions that can be used to perform an almost infinite number of calculations. This friendly book shows you how to use the 150 most valuable ones in real-world situations: to compare the cost of buying vs. leasing a car, calculate classroom grades, or evaluate investment performance, for example. Another 85 specialized functions are also described. Detailed, step-by-step instructions help you understand how functions work within formulas and how you can use them t

  12. Relationship between changes in the crystal lattice strain and thermal conductivity of high burnup UO{sub 2} pellets

    Energy Technology Data Exchange (ETDEWEB)

    Amaya, Masaki, E-mail: amaya.masaki@jaea.go.j [Fuel Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Nakamura, Jinichi; Fuketa, Toyoshi [Fuel Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kosaka, Yuji [Nuclear Development Corporation, 622-12, Funaishikawa, Tokai-mura, Naka-gun, Ibaraki 319-1111 (Japan)

    2010-01-01

    Two kinds of disk-shaped UO{sub 2} samples (4 mm in diameter and 1 mm in thickness) were irradiated in a test reactor up to about 60 and 130 GWd/t, respectively. The microstructures of the samples were investigated by means of optical microscopy, scanning electron microscopy/ electron probe micro-analysis (SEM/EPMA) and micro-X-ray diffractometry. The measured lattice parameters tended to be considerably smaller than the reported values, and the typical cauliflower structure which is often observed in high burnup fuel pellet is hardly seen in these samples. Thermal diffusivities of the samples were also measured by using a laser flash method, and their thermal conductivities were evaluated by multiplying the heat capacity of unirradiated UO{sub 2} and sample densities. While the thermal conductivities of sample 2 showed recovery after being annealed at 1500 K, those of sample 4 were not clearly observed even after being annealed at 1500 K. These trends suggest that the amount of accumulated irradiation-induced defects depends on the irradiation condition of each sample. From the comparison of the changes in the lattice parameter and strain energy density before and after the thermal diffusivity measurements, it is likely that the thermal conductivity recovery in the temperature region from 1200 to 1500 K is related to the migration of dislocation.

  13. Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

    Science.gov (United States)

    Bernstein, Adam; Bowden, Nathaniel S.; Erickson, Anna S.

    2018-01-01

    In a conventional light-water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under an assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for low-enriched uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields in the emitted antineutrino spectra and the detector response function, we show that the difference in corewide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

  14. Analysis of fresh fuel critical experiments appropriate for burnup credit validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Bowman, S.M.

    1995-10-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  15. Affine Grassmann codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Beelen, Peter; Ghorpade, Sudhir Ramakant

    2010-01-01

    We consider a new class of linear codes, called affine Grassmann codes. These can be viewed as a variant of generalized Reed-Muller codes and are closely related to Grassmann codes.We determine the length, dimension, and the minimum distance of any affine Grassmann code. Moreover, we show...

  16. School Leadership: Handbook for Excellence. Second Edition.

    Science.gov (United States)

    Smith, Stuart C., Ed.; Piele, Philip K., Ed.

    Recognizing the importance of good leadership to the achievement of educational excellence, the second edition of this handbook synthesizes a large body of school leadership literature and explores the subject from three perspectives: the person, the structure, and the skills. Part I examines characteristics of today's educational leaders;…

  17. Center of Excellence for Hypersonics Research

    Science.gov (United States)

    2012-01-25

    American astronauts on the moon . These observations motivated the need for a Center of Excellence in the area of hypersonic flows to advance the state of...as well as argon-xenon and argon- helium mixtures. The data produced by such first-principles MD simulations is then used to improve DSMC collision

  18. [Bodily excellence and education: ambiguities of performance].

    Science.gov (United States)

    Queval, Isabelle

    2011-01-01

    The notion of excellence contains an ambivalence: to aim at the "good" and/or to aim at the "best". This ambivalence exists particularly in the physical effort and in the history of bodily practices, shared between gymnastics, physical education and sports. In the second part of the XXth century, the notion of performance became "worship of performance", "infinite perfectibility". In this context, high level sports became the human improvement laboratory, thanks to the sophistication of its technical means and the its practices intensity. However this "high" questions: about physical and psychological consequences of intensive practices ; about doping and its medical and ethical perspectives; about the increasing precociousness and the potential exploitation of the baby champions. It questions about the nature of familial, social, economical norms transmitted by education: about articulation between constraint and self-government. This is the question about the "price" of excellence. In this sense it appears that the excellence of champion spreads out the margins of an ethics which would be meant to be regulating and universal, indeed this excellence is perhaps, in the same capacity as genius is and in spite of the attachment of sports to a "ethics of the rule", un-ethical.

  19. Using Excel in Teacher Education for Sustainability

    Science.gov (United States)

    Aydin, Serhat

    2016-01-01

    In this study, the feasibility of using Excel software in teaching whole Basic Statistics Course and its influence on the attitudes of pre-service science teachers towards statistics were investigated. One hundred and two pre-service science teachers in their second year participated in the study. The data were collected from the prospective…

  20. Introduction to 3D Graphics through Excel

    Science.gov (United States)

    Benacka, Jan

    2013-01-01

    The article presents a method of explaining the principles of 3D graphics through making a revolvable and sizable orthographic parallel projection of cuboid in Excel. No programming is used. The method was tried in fourteen 90 minute lessons with 181 participants, which were Informatics teachers, undergraduates of Applied Informatics and gymnasium…

  1. Schistosomacidal activity of Maytenus senegalensis (Lam.) Excell ...

    African Journals Online (AJOL)

    A study was conducted to investigate the schistosomacidal effect of Maytenus senegalensis (Lam.) Excell. on Schistosoma mansoni in mice. Three different concentrations (6.0, 3.0 and 1.5) mg/ml of the aqueous and ethanolic extracts of the plant roots was given orally to different groups of infected mice. Praziquantel was ...

  2. Excellence: The Key to Educating Visual Artists.

    Science.gov (United States)

    Rush, Jean C.; Conant, Howard

    1979-01-01

    The 1964-66 art education conferences were the last serious attempt to involve professional artists and educators in arts policy planning. This is a call to the art education profession to end this communication gap through a conference to examine the nature of excellence in the education of visual artists. (Author/SJL)

  3. Tips for Using Microsoft Excel in Assessment.

    Science.gov (United States)

    Kahan, David

    Acknowledging the need for additional practical resources for integrating technology into physical education assessment, this paper provides strategies and step-by-step instructions for incorporating Microsoft Excel into physical educators' assessment tools. The first section briefly reviews the literature on assessment in physical education. The…

  4. GUNSHOT FRACTURES OF TIBIA AND FEMUR - EXCELLENT ...

    African Journals Online (AJOL)

    2011-10-10

    Oct 10, 2011 ... GUNSHOT FRACTURES OF TIBIA AND FEMUR - EXCELLENT RESULTS WITH REAMED BONE MARROW GRAFT. AND INTERLOCKING .... between injury and surgery as well as associated injuries were documented. ..... injection in the treatment. of delayed and non-union in long bones. Singapore Med ...

  5. Expanding Educational Excellence: The Power of Schools

    Science.gov (United States)

    Coleman, Mary Ruth; Winn, Donna-Marie; Harradine, Christine

    2012-01-01

    In this paper, the authors explore four major barriers to academic success that must be addressed, briefly describe two projects that have worked to address these barriers, and make recommendations for moving forward as they work to expand educational excellence for all students. They provide examples of the myriad ways in which schools have the…

  6. Designing Quality Service: The Service Excellence Model.

    Science.gov (United States)

    Ellicott, Michael A.; Conard, Rodney J.

    1997-01-01

    Recent experiences of manufacturing and commercial service industries provide insights to college facilities managers for combining downsizing with quality improvement. The Service Excellence Model emphasizes creation of shared responsibility, focus on core service processes, empowerment of cross-functional process-improvement teams, performance…

  7. Humour: An Excellent EFL Teaching Device.

    Science.gov (United States)

    Guindal, Albert Lopez

    Humor is an excellent teaching tool because, in addition to preventing classroom boredom and monotony, it introduces lateral aspects of language such as irony, sarcasm, mockery, elision, ellipsis, and euphemism. Humor in language can be approached interactively or structurally through a variety of activities. It can be used to expand vocabulary,…

  8. MAINTAINING EXCELLENCE IN TEACHING OF HUMAN ANATOMY ...

    African Journals Online (AJOL)

    Kevin Wangwe Ongeti

    2013-01-02

    Jan 2, 2013 ... have resulted in wide disparities in curriculum design teaching methods, number and composition of instructors. Inspite of the challenges, the Department of Human Anatomy of the University of Nairobi. (UON) maintained excellence of teaching for over 40yrs. This article describes the teaching of anatomy.

  9. Do you write secure code?

    CERN Multimedia

    Computer Security Team

    2011-01-01

    At CERN, we are excellent at producing software, such as complex analysis jobs, sophisticated control programs, extensive monitoring tools, interactive web applications, etc. This software is usually highly functional, and fulfils the needs and requirements as defined by its author. However, due to time constraints or unintentional ignorance, security aspects are often neglected. Subsequently, it was even more embarrassing for the author to find out that his code flawed and was used to break into CERN computers, web pages or to steal data…   Thus, if you have the pleasure or task of producing software applications, take some time before and familiarize yourself with good programming practices. They should not only prevent basic security flaws in your code, but also improve its readability, maintainability and efficiency. Basic rules for good programming, as well as essential books on proper software development, can be found in the section for software developers on our security we...

  10. Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

    2014-10-15

    The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

  11. Does excellence matter? The influence of potential for excellence on students’ motivation for specific collaborative tasks

    NARCIS (Netherlands)

    A.G. Pullen (Annedien G.); D.M.E. Griffioen (Didi M. E.); J. Schoonenboom (Judith); B.B. de Koning (Björn); J.J. Beishuizen (Jos J.)

    2017-01-01

    textabstractIt is often assumed that students with a higher potential for excellence are less motivated to collaborate. So far, the question remains whether this is actually the case. This survey study investigated the influence of business students’ (N = 389) potential for excellence on their

  12. Turbo Codes Extended with Outer BCH Code

    DEFF Research Database (Denmark)

    Andersen, Jakob Dahl

    1996-01-01

    The "error floor" observed in several simulations with the turbo codes is verified by calculation of an upper bound to the bit error rate for the ensemble of all interleavers. Also an easy way to calculate the weight enumerator used in this bound is presented. An extended coding scheme is propose...... including an outer BCH code correcting a few bit errors.......The "error floor" observed in several simulations with the turbo codes is verified by calculation of an upper bound to the bit error rate for the ensemble of all interleavers. Also an easy way to calculate the weight enumerator used in this bound is presented. An extended coding scheme is proposed...

  13. Rateless feedback codes

    DEFF Research Database (Denmark)

    Sørensen, Jesper Hemming; Koike-Akino, Toshiaki; Orlik, Philip

    2012-01-01

    This paper proposes a concept called rateless feedback coding. We redesign the existing LT and Raptor codes, by introducing new degree distributions for the case when a few feedback opportunities are available. We show that incorporating feedback to LT codes can significantly decrease both...... the coding overhead and the encoding/decoding complexity. Moreover, we show that, at the price of a slight increase in the coding overhead, linear complexity is achieved with Raptor feedback coding....

  14. Generalized concatenated quantum codes

    Science.gov (United States)

    Grassl, Markus; Shor, Peter; Smith, Graeme; Smolin, John; Zeng, Bei

    2009-05-01

    We discuss the concept of generalized concatenated quantum codes. This generalized concatenation method provides a systematical way for constructing good quantum codes, both stabilizer codes and nonadditive codes. Using this method, we construct families of single-error-correcting nonadditive quantum codes, in both binary and nonbinary cases, which not only outperform any stabilizer codes for finite block length but also asymptotically meet the quantum Hamming bound for large block length.

  15. 1DB, a one-dimensional diffusion code for nuclear reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Little, W.W. Jr. (Pacific Northwest Lab., Richland, WA (United States))

    1991-09-01

    1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k{sub eff} and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor format; and to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs.

  16. TCODE: a computer code for analysis of tritium and vacuum systems for tokamak fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Clemmer, R.G.

    1978-08-01

    TCODE can be used for either near-term experimental reactors or for commercial reactors. The code provides options for items that may be included in a commercial reactor such as a divertor, neutral beam heating, and a breeding blanket. The code was used to calculate tritium and vacuum system parameters for the near term reactors ITR, TNS-UP and EPR as well as for some commercial reactor designs, the UWMAK series. A selected sample of the tritium and vacuum parameters for these reactor designs is shown. Also shown are parameters for a hypothetical reactor UWMAK-III M having similar characteristics to UWMAK-III but with a higher fractional burnup (5.0% cf. 0.83%). The impact of the reactor design scenario upon major tritium and vacuum systems is discussed.

  17. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

  18. Coding for dummies

    CERN Document Server

    Abraham, Nikhil

    2015-01-01

    Hands-on exercises help you learn to code like a pro No coding experience is required for Coding For Dummies,your one-stop guide to building a foundation of knowledge inwriting computer code for web, application, and softwaredevelopment. It doesn't matter if you've dabbled in coding or neverwritten a line of code, this book guides you through the basics.Using foundational web development languages like HTML, CSS, andJavaScript, it explains in plain English how coding works and whyit's needed. Online exercises developed by Codecademy, a leading online codetraining site, help hone coding skill

  19. Samuel P. Massie Chair of Excellence Program

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, James H. [Howard Univ., Washington, DC (United States)

    2014-12-01

    Abstract In 1994 the Department of Energy established the DOE Chair of Excellence Professorship in Environmental Disciplines Program. In 2004, the Massie Chair of Excellence Professor at Howard University transitioned from Dr. Edward Martin to Dr. James H. Johnson, Jr. At the time of his appointment Dr. Johnson served as professor of civil engineering and Dean of the College of Engineering, Architecture and Computer Sciences. Program activities under Dr. Johnson were in the following areas: • Increase the institution’s capacity to conduct scientific research and technical investigations at the cutting-edge. • Promote interactions, collaborations and partnerships between the private sector, Federal agencies, majority research institutes and other HBCUs. • Assist other HBCUs in reaching parity in engineering and related fields. • Mentor young investigators and be a role model for students.

  20. The Business Excellence Model for CSR Implementation?

    Directory of Open Access Journals (Sweden)

    Neergaard Peter

    2014-11-01

    Full Text Available Most of the Fortune 500 companies address Corporate Social Responsibility (CSR on their websites. However, CSR remains a fluffy concept difficult to implement in organization. The European Business Excellence Model has since the introduction in 1992 served as a powerful tool for integrating quality in organizations. CSR was first introduced in the model in 2002. From 2004 the European Foundation for Quality Management (EFQM has been eager to promote the model as an effective tool for implementing CSR.. The article discusses the potentials of the model for this end and illustrates how a 2006 European Award winning company has used the model to integrate CSR. The company adapted the Business Excellence model to improve performance, stimulate innovation and consensus.

  1. An EXCEL macro for importing log ASCII standard (LAS) files into EXCEL worksheets

    Science.gov (United States)

    Özkaya, Sait Ismail

    1996-02-01

    An EXCEL 5.0 macro is presented for converting a LAS text file into an EXCEL worksheet. Although EXCEL has commands for importing text files and parsing text lines, LAS files must be decoded line-by-line because three different delimiters are used to separate fields of differing length. The macro is intended to eliminate manual decoding of LAS version 2.0. LAS is a floppy disk format for storage and transfer of log data as text files. LAS was proposed by the Canadian Well Logging Society. The present EXCEL macro decodes different sections of a LAS file, separates, and places the fields into different columns of an EXCEL worksheet. To import a LAS file into EXCEL without errors, the file must not contain any unrecognized symbols, and the data section must be the last section. The program does not check for the presence of mandatory sections or fields as required by LAS rules. Once a file is incorporated into EXCEL, mandatory sections and fields may be inspected visually.

  2. Basic statistics with Microsoft Excel: a review.

    Science.gov (United States)

    Divisi, Duilio; Di Leonardo, Gabriella; Zaccagna, Gino; Crisci, Roberto

    2017-06-01

    The scientific world is enriched daily with new knowledge, due to new technologies and continuous discoveries. The mathematical functions explain the statistical concepts particularly those of mean, median and mode along with those of frequency and frequency distribution associated to histograms and graphical representations, determining elaborative processes on the basis of the spreadsheet operations. The aim of the study is to highlight the mathematical basis of statistical models that regulate the operation of spreadsheets in Microsoft Excel.

  3. Independent School Libraries Perspectives on excellence

    CERN Document Server

    HAND, DORCAS

    2010-01-01

    Independent School Libraries: Perspectives on Excellence offers readers insights into best practices in library services for school communities, using examples drawn from independent schools of various sizes, descriptions, and locations across the United States. Two overview essays introduce a statistical analysis of independent schools. Each of the remaining essays provides perspective on a different aspect of library practice, including staffing, advocacy, assessment, technology, collaboration, programs beyond the curriculum, intellectual freedom and privacy, budgeting, accreditation, disast

  4. Living excellence: life after Magnet designation.

    Science.gov (United States)

    Malloch, Kathy

    2009-01-01

    The achievement of Magnet recognition is the beginning of a new way of being as an organization. Strategies to support innovation leadership, value-based decision making, agility, sustainability of excellence, technology advancements, and lifelong learning are discussed within the framework of the Magnet organization. Behaviors and challenges of living the expectations of the Magnet organization are presented as opportunities to assist healthcare leaders in this important work.

  5. Forecasting with Excel: Suggestions for Managers

    Directory of Open Access Journals (Sweden)

    Scott Nadler

    2007-05-01

    Full Text Available This article provides readers and more especially business managers with an overview of moving average, exponential smoothing, trend analysis, and linear regression approaches to forecasting. The authors then provide specific examples for each approach and the Excel formulas necessary to develop effective forecasts. This is an important contribution to the literature because it demonstrates that businesses with limited resources can develop reliable and accurate forecasts in a timely and cost effective manner using readily available software.

  6. Separation of metallic residues from the dissolution of a high-burnup BWR fuel using nitrogen trifluoride

    Energy Technology Data Exchange (ETDEWEB)

    McNamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Scheele, Randall D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-03-23

    Nitrogen trifluoride (NF3) was used to fluorinate the metallic residue from the dissolution of a high burnup, boiling water reactor fuel (~70 MWd/kgU). The metallic residue included the noble metal phase (containing ruthenium, rhodium, palladium, technetium, and molybdenum), and smaller amounts of zirconium, selenium, tellurium, and silver. Exposing the noble metal phase to 10% NF3 in argon between 400 and 550°C, removed molybdenum and technetium near 400°C as their volatile fluorides, and ruthenium near 500C as its volatile fluoride. The events were thermally and temporally distinct and the conditions specified are a recipe to separate these transition metals from each other and from the noble metal phase nonvolatile residue. Depletion of the volatile fluorides resulted in substantial exothermicity. Thermal excursion behavior was recorded under non-adiabatic, isothermal conditions that typically minimize heat release. Physical characterization of the metallic noble phase and its thermal behavior are consistent with high kinetic velocity reactions encouraged by the nanoparticulate phase or perhaps catalytic influences of the mixed platinum metals with nearly pure phase structure. Post-fluorination, only two phases were present in the residual nonvolatile fraction. These were identified as a nano-crystalline, metallic palladium cubic phase and a hexagonal rhodium trifluoride (RhF3) phase. The two phases were distinct as the sub-µm crystallites of metallic palladium were in contrast to the RhF3 phase, which grew from the parent nano-crystalline noble-metal phase during fluorination, to acicular crystals exceeding 20-µm in length.

  7. Discussion on LDPC Codes and Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the progress that the workgroup on Low-Density Parity-Check (LDPC) for space link coding. The workgroup is tasked with developing and recommending new error correcting codes for near-Earth, Lunar, and deep space applications. Included in the presentation is a summary of the technical progress of the workgroup. Charts that show the LDPC decoder sensitivity to symbol scaling errors are reviewed, as well as a chart showing the performance of several frame synchronizer algorithms compared to that of some good codes and LDPC decoder tests at ESTL. Also reviewed is a study on Coding, Modulation, and Link Protocol (CMLP), and the recommended codes. A design for the Pseudo-Randomizer with LDPC Decoder and CRC is also reviewed. A chart that summarizes the three proposed coding systems is also presented.

  8. Locally orderless registration code

    DEFF Research Database (Denmark)

    2012-01-01

    This is code for the TPAMI paper "Locally Orderless Registration". The code requires intel threadding building blocks installed and is provided for 64 bit on mac, linux and windows.......This is code for the TPAMI paper "Locally Orderless Registration". The code requires intel threadding building blocks installed and is provided for 64 bit on mac, linux and windows....

  9. Data acquisition and real-time control using spreadsheets: interfacing Excel with external hardware.

    Science.gov (United States)

    Aliane, Nourdine

    2010-07-01

    Spreadsheets have become a popular computational tool and a powerful platform for performing engineering calculations. Moreover, spreadsheets include a macro language, which permits the inclusion of standard computer code in worksheets, and thereby enable developers to greatly extend spreadsheets' capabilities by designing specific add-ins. This paper describes how to use Excel spreadsheets in conjunction to Visual Basic for Application programming language to perform data acquisition and real-time control. Afterwards, the paper presents two Excel applications with interactive user interfaces developed for laboratory demonstrations and experiments in an introductory course in control. 2010 ISA. Published by Elsevier Ltd. All rights reserved.

  10. Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A. [and others

    1995-12-31

    In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.

  11. Algebraic geometric codes

    Science.gov (United States)

    Shahshahani, M.

    1991-01-01

    The performance characteristics are discussed of certain algebraic geometric codes. Algebraic geometric codes have good minimum distance properties. On many channels they outperform other comparable block codes; therefore, one would expect them eventually to replace some of the block codes used in communications systems. It is suggested that it is unlikely that they will become useful substitutes for the Reed-Solomon codes used by the Deep Space Network in the near future. However, they may be applicable to systems where the signal to noise ratio is sufficiently high so that block codes would be more suitable than convolutional or concatenated codes.

  12. Monomial-like codes

    CERN Document Server

    Martinez-Moro, Edgar; Ozbudak, Ferruh; Szabo, Steve

    2010-01-01

    As a generalization of cyclic codes of length p^s over F_{p^a}, we study n-dimensional cyclic codes of length p^{s_1} X ... X p^{s_n} over F_{p^a} generated by a single "monomial". Namely, we study multi-variable cyclic codes of the form in F_{p^a}[x_1...x_n] / . We call such codes monomial-like codes. We show that these codes arise from the product of certain single variable codes and we determine their minimum Hamming distance. We determine the dual of monomial-like codes yielding a parity check matrix. We also present an alternative way of constructing a parity check matrix using the Hasse derivative. We study the weight hierarchy of certain monomial like codes. We simplify an expression that gives us the weight hierarchy of these codes.

  13. Excellence in the knowledge-based economy: from scientific to research excellence

    DEFF Research Database (Denmark)

    Sørensen, Mads P.; Bloch, Carter Walter; Young, Mitchell

    2016-01-01

    In 2013, the European Union (EU) unveiled its new ‘Composite Indicator for Scientific and Technological Research Excellence’. This is not an isolated occurrence; policy-based interest in excellence is growing all over the world. The heightened focus on excellence and, in particular, attempts....... This change is evidenced by the ‘Composite Indicator for Scientific and Technological Research Excellence’, its rationale and its components....

  14. QR Codes 101

    Science.gov (United States)

    Crompton, Helen; LaFrance, Jason; van 't Hooft, Mark

    2012-01-01

    A QR (quick-response) code is a two-dimensional scannable code, similar in function to a traditional bar code that one might find on a product at the supermarket. The main difference between the two is that, while a traditional bar code can hold a maximum of only 20 digits, a QR code can hold up to 7,089 characters, so it can contain much more…

  15. Determination of Fission Gas Inclusion Pressures in High Burnup Nuclear Fuel using Laser Ablation ICP-MS combined with SEM/EPMA and Optical Microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Horvath, Matthias I.; Guenther-Leopold, Ines; Kivel, Niko; Restani, Renato [Laboratory for Materials Behavior, Nuclear Energy and Safety, Paul Scherrer Institut, Villigen, CH-5232 (Switzerland); Guillong, Marcel [Institute for Isotope Geology/Mineralogic Elements, ETH Zuerich, CH-8092 (Switzerland); Izmer, Andrei [Environmental and Resource Studies, Trent University, Peterborough, K9J 7B8 (Canada); Hellwig, Christian [Nuclear Technology Department, Nordostschweizerische Kraftwerke AG (NOK), Baden, CH-5401 (Switzerland); Guenther, Detlef [Laboratory for Inorganic Chemistry, Trace Elements and Microanalysis Group, ETH Zuerich, CH-8093 (Switzerland)

    2008-07-01

    In approximately 20% of all fissions at least one of the fission products is gaseous. These are mainly xenon and krypton isotopes contributing up to 90% by the xenon isotopes. Upon reaching a burn-up of 60 - 75 GWd/tHM a so called High Burnup Structure (HBS) is formed in the cooler rim of the fuel. In this region a depletion of the noble fission gases (FG) in the matrix and an enrichment of FG in {mu}m-sized pores can be observed. Recent calculations show that in these pores the pressure at room temperature can be as large as 30 MPa. The knowledge of the FG pressure in pores is important to understand the high burn-up fuel behavior under accident conditions (i.e. RIA or LOCA). With analytical methods routinely used for the characterization of solid samples, i.e. Electron Probe Micro Analysis (EPMA), Secondary Ion Mass Spectrometry (SIMS), the quantification of gaseous inclusions is very difficult to almost impossible. The combination of a laser ablation system (LA) with an inductively coupled plasma mass spectrometer (ICP-MS) offers a powerful tool for quantification of the gaseous pore inventory. This method offers the advantages of high spatial resolution with laser spot sizes down to 10 {mu}m and low detection limits. By coupling with scanning electron microscopy (SEM) for the pore size distribution, EPMA for the FG inventory in the fuel matrix and optical microscopy for the LA-crater sizes, the pressures in the pores and porosity was calculated. As a first application of this calibration technique for gases, measurements were performed on pressurized water reactor (PWR) fuel with a rod average of 105 GWd/tHM to determine the local FG pressure distribution. (authors)

  16. Modeling and Analysis of FCM UN TRISO Fuel Using the PARFUME Code

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Collin

    2013-09-01

    The PARFUME (PARticle Fuel ModEl) modeling code was used to assess the overall fuel performance of uranium nitride (UN) tri-structural isotropic (TRISO) ceramic fuel in the frame of the design and development of Fully Ceramic Matrix (FCM) fuel. A specific modeling of a TRISO particle with UN kernel was developed with PARFUME, and its behavior was assessed in irradiation conditions typical of a Light Water Reactor (LWR). The calculations were used to access the dimensional changes of the fuel particle layers and kernel, including the formation of an internal gap. The survivability of the UN TRISO particle was estimated depending on the strain behavior of the constituent materials at high fast fluence and burn-up. For nominal cases, internal gas pressure and representative thermal profiles across the kernel and layers were determined along with stress levels in the pyrolytic carbon (PyC) and silicon carbide (SiC) layers. These parameters were then used to evaluate fuel particle failure probabilities. Results of the study show that the survivability of UN TRISO fuel under LWR irradiation conditions might only be guaranteed if the kernel and PyC swelling rates are limited at high fast fluence and burn-up. These material properties are unknown at the irradiation levels expected to be reached by UN TRISO fuel in LWRs. Therefore, more effort is needed to determine them and positively conclude on the applicability of FCM fuel to LWRs.

  17. Preliminary Core Analysis of High Temperature Engineering Test Reactor Using DeCART Code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The 2-dimensional core analysis for the High Temperature Engineering Test Reactor (HTTR) has been performed. The HTTR is a graphite-moderated and helium gas cooled reactor with an outlet temperature of 950 .deg. C and thermal output of 30 MW. In this study, the DECART code is used with a 190-group KARMA library. The calculation results are compared with those of the McCARD with the ENDF-B/VII.0 library. From the analysis results, it is known that the DeCART code generally overestimates k{sub inf} with a moderator temperature variation. In addition, it can be seen that the DeCART code predicts less negative MTC than the McCARD code. However, the DeCART code gives a slightly more negative FTC value. From the depletion results, the error of the DeCART decreases over the burnup until 600 FPD. The DeCART code gives very similar trend within the error of 190 pcm, which is very small error when compared with other result.

  18. V.S.O.P. (99/05) computer code system

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Scherer, W.

    2005-11-01

    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 HTRs and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. V.S.O.P.(99 / 05) represents the further development of V.S.O.P. (99). Compared to its precursor, the code system has been improved in many details. Major improvements and extensions have been included concerning the neutron spectrum calculation, the 3-d neutron diffusion options, and the thermal hydraulic section with respect to 'multi-pass'-fuelled pebblebed cores. This latest code version was developed and tested under the WINDOWS-XP - operating system. The storage requirement for the executables and the basic libraries associated with the code amounts to about 15 MB. Another 5 MB are required - if desired - for storage of the source code ({approx}65000 Fortran statements). (orig.)

  19. Tutorial: simulating chromatography with Microsoft Excel Macros.

    Science.gov (United States)

    Kadjo, Akinde; Dasgupta, Purnendu K

    2013-04-22

    Chromatography is one of the cornerstones of modern analytical chemistry; developing an instinctive feeling for how chromatography works will be invaluable to future generation of chromatographers. Specialized software programs exist that handle and manipulate chromatographic data; there are also some that simulate chromatograms. However, the algorithm details of such software are not transparent to a beginner. In contrast, how spreadsheet tools like Microsoft Excel™ work is well understood and the software is nearly universally available. We show that the simple repetition of an equilibration process at each plate (a spreadsheet row) followed by discrete movement of the mobile phase down by a row, easily automated by a subroutine (a "Macro" in Excel), readily simulates chromatography. The process is readily understood by a novice. Not only does this permit simulation of isocratic and simple single step gradient elution, linear or multistep gradients are also easily simulated. The versatility of a transparent and easily understandable computational platform further enables the simulation of complex but commonly encountered chromatographic scenarios such as the effects of nonlinear isotherms, active sites, column overloading, on-column analyte degradation, etc. These are not as easily simulated by available software. Views of the separation as it develops on the column and as it is seen by an end-column detector are both available in real time. Excel 2010™ also permits a 16-level (4-bit) color gradation of numerical values in a column/row; this permits visualization of a band migrating down the column, much as Tswett may have originally observed, but in a numerical domain. All parameters of relevance (partition constants, elution conditions, etc.) are readily changed so their effects can be examined. Illustrative Excel spreadsheets are given in the Supporting Information; these are easily modified by the user or the user can write his/her own routine. Copyright

  20. SAS6. User`s guide. A two-dimensional depletion and criticality analysis code package based on the SCALE-4 system

    Energy Technology Data Exchange (ETDEWEB)

    Leege, P.F.A. de; Li, J.M.; Kloosterman, J.L.

    1995-04-01

    This users` guide gives a description of the functionality and the requested input of the SAS6 code sequence which can be used to perform burnup and criticality calculations based on functional modules from the SCALE-4 code system and libraries. The input file for the SAS6 control module is very similar to that of the other SAS and CSAS control modules available in the SCALE-4 system. Especially the geometry input of SAS6 is quite similar to that of SAS2H. However, the functionality of SAS6 is different from that of SAS2H. The geometry of the reactor lattice can be treated in more detail because use is made of the two-dimensional lattice code WIMS-D/IRI (An adapted version of WIMS-D/4) instead of the one-dimensional transport code XSDRNPM-S. Also the neutron absorption and production rates of nuclides not explicitly specified in the input can be accounted for by six pseudo nuclides. Furthermore, the centre pin can be subdivided into maximal 10 zones to improve the burnup calculation of the centre pin and to obtain more accurate k-infinite values for the assembly. Also the time step specification is more flexible than in the SAS2H sequence. (orig.).

  1. Using Excel to reduce a Square Matrix

    Directory of Open Access Journals (Sweden)

    Josef Holoubek

    2012-01-01

    Full Text Available When solving operations research problems, one can use either specialised computer programs such as Lingo, Lindo, Storm or more universal programs such Excel, Matlab, and R. To obtain the input data, one can use either a program’s own editor or other programs commonly available such as Excel. While the problem-solving methods, being part of various programs, are the subjects of numerous publications (such as Gros, 2003; Jablonský, 2002; Plevný – Žižka, 2007; Stevenson – Ozgur, 2009, the way the input data are obtained, recorded, and processed receives far less attention although this part of problem-solving requires considerable effort and, if the method for data recording is inadequate, may cause subsequent difficulties in their further processing. A problem known as “the travelling salesman problem” (TSP may serve as an example. Here, the input data form a “square matrix of distances”. This paper is concerned with some Excel tools that can be used to obtain and subsequently modify such a square matrix. Given a square m × m matrix, an ordinary user might want to reduce it to an i × i square matrix (where i < m without having to copy data from the matrix, skip some of its rows and/or columns or write a program to implement such a reduction.In her degree project, Kourková, 2009 was looking for an efficient method of reducing an Excel matrix. She had found no relevant papers on this subject concluding that the authors of the commercial program had not considered this. Therefore, she offered her own solution unconventionally using the contingency table menu option. Although this had resulted in the desired submatrix, some of its parts were superfluous and even baffling for the user.For this reason, the authors analyse the method of representing an m × m matrix and the way of its reduction. Finally, a better option is offered to achieve the desired objective as well as other methods of obtaining the required submatrix that even

  2. Graphing techniques for materials laboratory using Excel

    Science.gov (United States)

    Kundu, Nikhil K.

    1994-01-01

    Engineering technology curricula stress hands on training and laboratory practices in most of the technical courses. Laboratory reports should include analytical as well as graphical evaluation of experimental data. Experience shows that many students neither have the mathematical background nor the expertise for graphing. This paper briefly describes the procedure and data obtained from a number of experiments such as spring rate, stress concentration, endurance limit, and column buckling for a variety of materials. Then with a brief introduction to Microsoft Excel the author explains the techniques used for linear regression and logarithmic graphing.

  3. Internationalization as Universities’ Strive for Excellence

    Directory of Open Access Journals (Sweden)

    Nina Bijedic

    2015-07-01

    Full Text Available Internationalization, when not observed in the narrow sense of mare mobility, assumes all processes that lead to international recognition of a university. Consequently, one can think of internationalization as of all supporting processes in university’s strive for widely recognized excellence. In order to use internationalization as a drive for changes, it is necessary to observe it through different aspects: strategic documents, teaching process, RDI policy, international projects, mobility programs and quality assurance. A useful tool for monitoring success of planed processes and measures is internal benchmarking, used to bring the institution ever closer to strategic goals.

  4. Le choix de l'excellence

    CERN Document Server

    Collins, Jim

    2012-01-01

    Dix ans après le succès mondial de De la performance à l'excellence (Good to Great), Jim Collins et son associé Morten T. Hansen s'interrogent ici sur la nature de l excellence en période d incertitude. Pourquoi, dans un contexte chaotique, certaines entreprises sont florissantes et d autres non ? Comment, en ces temps troublés, réussir à diriger et vaincre l adversité pour dominer son marché ? La chance peut-elle sérieusement être envisagée comme un facteur de réussite ? À l issue d un programme de recherche ambitieux et rigoureux qui a duré neuf ans, les auteurs exposent dans cet ouvrage les nouveaux principes pour bâtir une entreprise réellement excellente. Leurs conclusions sont, comme toujours, étonnantes. Ils soulignent notamment que la course à l innovation n est pas toujours la meilleure voie à suivre, que discipline et créativité peuvent aller de concert et que malgré le changement permanent, les meilleures décisions sont rarement prises dans l urgence. Quant aux leaders, on...

  5. Discrimination against Women as a Subtext of Excellence.

    Science.gov (United States)

    Watson, Scott B.

    1993-01-01

    Pursuing excellence in sport discriminates against women. Alongside the primary texts of the glories of excellence runs a subtext of inequality and discrimination. The paper traces the idea of sport as a quest for excellence through various interpretations of the meaning of sport, discussing excellence as a socially constructed concept. (SM)

  6. A simple program to measure and analyse tree rings using Excel, R and SigmaScan

    Science.gov (United States)

    Hietz, Peter

    2011-01-01

    I present a new software that links a program for image analysis (SigmaScan), one for spreadsheets (Excel) and one for statistical analysis (R) for applications of tree-ring analysis. The first macro measures ring width marked by the user on scanned images, stores raw and detrended data in Excel and calculates the distance to the pith and inter-series correlations. A second macro measures darkness along a defined path to identify latewood–earlywood transition in conifers, and a third shows the potential for automatic detection of boundaries. Written in Visual Basic for Applications, the code makes use of the advantages of existing programs and is consequently very economic and relatively simple to adjust to the requirements of specific projects or to expand making use of already available code. PMID:26109835

  7. A simple program to measure and analyse tree rings using Excel, R and SigmaScan.

    Science.gov (United States)

    Hietz, Peter

    I present a new software that links a program for image analysis (SigmaScan), one for spreadsheets (Excel) and one for statistical analysis (R) for applications of tree-ring analysis. The first macro measures ring width marked by the user on scanned images, stores raw and detrended data in Excel and calculates the distance to the pith and inter-series correlations. A second macro measures darkness along a defined path to identify latewood-earlywood transition in conifers, and a third shows the potential for automatic detection of boundaries. Written in Visual Basic for Applications, the code makes use of the advantages of existing programs and is consequently very economic and relatively simple to adjust to the requirements of specific projects or to expand making use of already available code.

  8. John Walkenbach's Favorite Excel 2010 Tips and Tricks

    CERN Document Server

    Walkenbach, John

    2010-01-01

    Build robust Excel 2010 apps quickly and efficiently. Known as "Mr. Spreadsheet," John Walkenbach's name is synonymous with excellence in computer books that explain the complexities of various topics. With this collection of favorite Excel tips and tricks, you get a unique look at ways to make Excel 2010 more efficient than you ever imagined. Packed with easy-to-understand advice regarding all aspects of Excel, this book shares improved ways of speeding up application development with Excel and maximizing the power of Excel to create robust applications.: Addresses the extensive cha

  9. Simplifying CEA through Excel, VBA, and Subeq

    Science.gov (United States)

    Foster, Ryan

    2004-01-01

    Many people use compound equilibrium programs for very different reasons, varying from refrigerators to light bulbs to rockets. A commonly used equilibrium program is CEA. CEA can take various inputs such as pressure, temperature, and volume along with numerous reactants and run them through equilibrium equations to obtain valuable output information, including products formed and their relative amounts. A little over a year ago, Bonnie McBride created the program subeq with the goal to simplify the calling of CEA. Subeq was also designed to be called by other programs, including Excel, through the use of Visual Basic for Applications (VBA). The largest advantage of using Excel is that it allows the user to input the information in a colorful and user-friendly environment while allowing VBA to run subeq, which is in the form of a FORTRAN DLL (Dynamic Link Library). Calling subeq in this form makes it much faster than if it were converted to VBA. Since subeq requires such large lists of reactant and product names, all of which can't be passed in as an array, subeq had to be changed to accept very long strings of reactants and products. To pass this string and adjust the transfer of input and output parameters, the subeq DLL had to be changed. One program that does this is Compaq Visual FORTRAN, which allows DLLs to be edited, debugged, and compiled. Compaq Visual FORTRAN uses FORTRAN 90/95, which has additional features to that of FORTRAN 77. My goals this summer include finishing up the excel spreadsheet of subeq, which I started last summer, and putting it on the Internet so that others can use it without having to download my spreadsheet. To finish up the spreadsheet I will need to work on debugging current options and problems. I will also work on making it as robust as possible, so that all errors that may arise will be clearly communicated to the user. New features will be added old ones will be changed as I receive comments from people using the spreadsheet

  10. Excellence in role modelling: insight and perspectives from the pros.

    Science.gov (United States)

    Wright, Scott M; Carrese, Joseph A

    2002-09-17

    Role modelling is an effective teaching method in medical education. We sought to better understand role modelling by examining the insights of respected physician role models. We conducted 30-minute in-depth interviews with 29 highly regarded role models at 2 large teaching hospitals. We coded the transcripts independently, and compared our coding for agreement. Content analysis identified several major categories of themes. The informants identified specific characteristics related to role modelling. Subcategories under the domain of personal qualities included interpersonal skills, a positive outlook, a commitment to excellence and growth, integrity and leadership. Under the domain of teaching, the subcategories were establishing rapport with learners, developing specific teaching philosophies and methods, and being committed to the growth of learners. Subjects thought there was some overlap between teaching and role modelling, but felt that the latter was more implicit and more encompassing. Being a strong clinician was regarded as necessary but not sufficient for being an exemplary physician role model. Perceived barriers to effective role modelling included being impatient and overly opinionated, being quiet, being overextended, and having difficulty remembering names and faces. Physician role models described role modeling consciousness, in that they specifically think about being role models when interacting with learners. Subjects believed that medical learners should emulate multiple role models. Highly regarded physician role models possess personal qualities, teaching abilities and exceptional clinical skills that outweigh their own barriers to serving as effective role models. Many of these positive attributes of role models represent behaviours that can be modified or skills that can be acquired.

  11. Guaranteeing CERN’s excellence: consolidate experience

    CERN Multimedia

    Staff Association

    2014-01-01

    For its missions CERN requires staff with solid experience in all its domains of activity The Organization has several missions: fundamental research, technical developments and innovation, training the several hundreds of associates, fellows and students, while at the same time taking care of more than 10000 users. In order to guarantee excellence in all of these areas, CERN has to put in place an efficient personnel policy. Such a policy must allow the Organization to recruit collaborators with the highest competence from all Member States, and to keep and motivate them during their entire professional career. But, more importantly, the Organization has to be able to count on a stable workforce. It needs staff with experience gained over a long period in the fields of accelerators, detectors, and operating procedures, if it is to fulfil successfully its important mission of training and knowledge transfer, which is very much appreciated by our Member states, since it highlights a visible return on invest...

  12. Pension fund excellence creating value for stakeholders

    CERN Document Server

    Ambachtsheer, Keith P.

    1998-01-01

    Internationally recognized experts in the field introduce their "business excellence paradigm". In this book, two leading pension fund experts lay out a comprehensive plan for effective fund management. With the help of domestic and global case studies they critically assess current approaches to pension fund management and isolate what works and what doesn't using their unique critically acclaimed "run-it-like-a-business" model. Keith P. Ambachtsheer (Toronto, Canada) is principle at KPA Advisory Service, Inc., a pension fund management consulting firm. He runs The Ambachtsheer Letter and cofounded Cost Effective Measurement, Inc., which monitors the performance of 300 of the world's largest asset funds. D. Don Ezra (Toronto, Canada) is Director of European Consulting at Frank Russell Co. His previous books include The Struggle for Pension Fund Wealth.

  13. VIP Visit Her Excellency Dr Dalia Grybauskaite

    CERN Multimedia

    2016-01-01

    Her Excellency Dr Dalia Grybauskaite President Republic of Lithuania. Wednesday 20 January 2016. General introduction to CERN’s activities by CERN Director-General F. Gianotti. CERN Director-General introduces CMS Collaboration Deputy Spokesperson K. Borras and CMS Lithuanians A. Rinkevicius with V. Rapsevicius. Met by J. Shiers, IT Department Data Preservation Project Leader, and walk to 1st floor Data Centre Visit Point. CERN Data Centre Visit Point, 1st floor, building 513. View the robotic arms of the CERN IT data centre automated libraries (J. Shiers) CERN Computer Centre, building 513, level -1. Physics hands-on and virtual visit with a High School class in Lithuania (S. Schmeling) CERN S’cool Lab, building 143. Meeting with the Lithuanian community at CERN. Signature of the Guest Book with CERN Director-General witnessed by the Lithuanian community at CERN. Family photograph with the Lithuanian community at CERN.

  14. A national neurological excellence centers network.

    Science.gov (United States)

    Pazzi, S; Cristiani, P; Cavallini, A

    1998-02-01

    The most relevant problems related to the management of neurological disorders are (i) the frequent hospitalization in nonspecialist departments, with the need for neurological consultation, and (ii) the frequent requests of GPs for highly specialized investigations that are very expensive and of little value in arriving at a correct diagnosis. In 1996, the Consorzio di Bioingegneria e Informatica Medica in Italy realized the CISNet project (in collaboration with the Consorzio Istituti Scientifici Neuroscienze e Tecnologie Biomediche and funded by the Centro Studi of the National Public Health Council) for the implementation of a national neurological excellence centers network (CISNet). In the CISNet project, neurologists will be able to give on-line interactive consultation and off-line consulting services identifying correct diagnostic/therapeutic procedures, evaluating the need for both examination in specialist centers and admission to specialized centers, and identifying the most appropriate ones.

  15. Business excellence journey in countries in transition

    Directory of Open Access Journals (Sweden)

    Pavel Castka

    2001-01-01

    Full Text Available This paper describes the journey toward business excellence regarding the influence of the contemporary global environment, as well as the specifics of the environment in countries in transition (the environment in the Czech Republic is taken as a representative model. Closer focus is given on problems connected with productivity, effectivity, innovation, quality and certification, use of IT/IS and on problems with participation of employees. The abovementioned characteristics make up the goals of a change to a process-oriented company. In regard of business downturn in many companies in the post-communist environment, these goals are achievable by radical change using the ideas of Business Process Reengineering (BPR and the process-oriented model described at the end of this paper.

  16. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  17. TIPONLINE Code Table

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Coded items are entered in the tiponline data entry program. The codes and their explanations are necessary in order to use the data

  18. Coding for optical channels

    CERN Document Server

    Djordjevic, Ivan; Vasic, Bane

    2010-01-01

    This unique book provides a coherent and comprehensive introduction to the fundamentals of optical communications, signal processing and coding for optical channels. It is the first to integrate the fundamentals of coding theory and optical communication.

  19. Performance Analysis for Cooperative Communication System with QC-LDPC Codes Constructed with Integer Sequences

    Directory of Open Access Journals (Sweden)

    Yan Zhang

    2015-01-01

    Full Text Available This paper presents four different integer sequences to construct quasi-cyclic low-density parity-check (QC-LDPC codes with mathematical theory. The paper introduces the procedure of the coding principle and coding. Four different integer sequences constructing QC-LDPC code are compared with LDPC codes by using PEG algorithm, array codes, and the Mackey codes, respectively. Then, the integer sequence QC-LDPC codes are used in coded cooperative communication. Simulation results show that the integer sequence constructed QC-LDPC codes are effective, and overall performance is better than that of other types of LDPC codes in the coded cooperative communication. The performance of Dayan integer sequence constructed QC-LDPC is the most excellent performance.

  20. ARC Code TI: ROC Curve Code Augmentation

    Data.gov (United States)

    National Aeronautics and Space Administration — ROC (Receiver Operating Characteristic) curve Code Augmentation was written by Rodney Martin and John Stutz at NASA Ames Research Center and is a modification of ROC...

  1. Gauge color codes

    DEFF Research Database (Denmark)

    Bombin Palomo, Hector

    2015-01-01

    Color codes are topological stabilizer codes with unusual transversality properties. Here I show that their group of transversal gates is optimal and only depends on the spatial dimension, not the local geometry. I also introduce a generalized, subsystem version of color codes. In 3D they allow...

  2. Refactoring test code

    NARCIS (Netherlands)

    A. van Deursen (Arie); L.M.F. Moonen (Leon); A. van den Bergh; G. Kok

    2001-01-01

    textabstractTwo key aspects of extreme programming (XP) are unit testing and merciless refactoring. Given the fact that the ideal test code / production code ratio approaches 1:1, it is not surprising that unit tests are being refactored. We found that refactoring test code is different from

  3. The Procions` code; Le code Procions

    Energy Technology Data Exchange (ETDEWEB)

    Deck, D.; Samba, G.

    1994-12-19

    This paper presents a new code to simulate plasmas generated by inertial confinement. This multi-kinds kinetic code is done with no angular approximation concerning ions and will work in plan and spherical geometry. First, the physical model is presented, using Fokker-Plank. Then, the numerical model is introduced in order to solve the Fokker-Plank operator under the Rosenbluth form. At the end, several numerical tests are proposed. (TEC). 17 refs., 27 figs.

  4. Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

    2007-07-15

    Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

  5. The materiality of Code

    DEFF Research Database (Denmark)

    Soon, Winnie

    2014-01-01

    , Twitter and Facebook). The focus is not to investigate the functionalities and efficiencies of the code, but to study and interpret the program level of code in order to trace the use of various technological methods such as third-party libraries and platforms’ interfaces. These are important...... to understand the socio-technical side of a changing network environment. Through the study of code, including but not limited to source code, technical specifications and other materials in relation to the artwork production, I would like to explore the materiality of code that goes beyond technical...

  6. Excel 2010 all-in-one for dummies

    CERN Document Server

    Harvey, Greg

    2010-01-01

    A comprehensive, up-to-date, user-friendly guide to Excel 2010 Excel is the standard for spreadsheet applications and is used worldwide, but it's not always user-friendly. That makes it a perfect For Dummies topic, and this handy all-in-one guide covers all the essentials, the new features, how to analyze data with Excel, and much more. Eight minibooks address Excel basics, worksheet design, formulas and functions, worksheet collaboration and review, charts and graphics, data management, data analysis, and Excel and VBA.Excel is the leading spreadsheet/data analysis s

  7. DLLExternalCode

    Energy Technology Data Exchange (ETDEWEB)

    2014-05-14

    DLLExternalCode is the a general dynamic-link library (DLL) interface for linking GoldSim (www.goldsim.com) with external codes. The overall concept is to use GoldSim as top level modeling software with interfaces to external codes for specific calculations. The DLLExternalCode DLL that performs the linking function is designed to take a list of code inputs from GoldSim, create an input file for the external application, run the external code, and return a list of outputs, read from files created by the external application, back to GoldSim. Instructions for creating the input file, running the external code, and reading the output are contained in an instructions file that is read and interpreted by the DLL.

  8. Noisy Network Coding

    CERN Document Server

    Lim, Sung Hoon; Gamal, Abbas El; Chung, Sae-Young

    2010-01-01

    A noisy network coding scheme for sending multiple sources over a general noisy network is presented. For multi-source multicast networks, the scheme naturally extends both network coding over noiseless networks by Ahlswede, Cai, Li, and Yeung, and compress-forward coding for the relay channel by Cover and El Gamal to general discrete memoryless and Gaussian networks. The scheme also recovers as special cases the results on coding for wireless relay networks and deterministic networks by Avestimehr, Diggavi, and Tse, and coding for wireless erasure networks by Dana, Gowaikar, Palanki, Hassibi, and Effros. The scheme involves message repetition coding, relay signal compression, and simultaneous decoding. Unlike previous compress--forward schemes, where independent messages are sent over multiple blocks, the same message is sent multiple times using independent codebooks as in the network coding scheme for cyclic networks. Furthermore, the relays do not use Wyner--Ziv binning as in previous compress-forward sch...

  9. CERN, a laboratory of social excellence

    CERN Multimedia

    Staff Association

    2012-01-01

    The HR 2012 Forum, where Human Resources (HR) professionals meet and present innovative services and tools in the HR sector, was held on 3rd and 4th October at Palexpo, Geneva. CERN won the 3rd HR Innovation Award for the Long-Term Saved Leave Scheme (LTSLS) for its novelty and innovation. The award was presented to Jean-Marc Saint Viteux, Deputy-Head of HR Department. This is excellent news for our Organization. One can say that the Long-Term Saved Leave Scheme (LTSLS) originated from proposals by the Staff Association in 1996, based on general considerations on a framework for flexible working time arrangements. This article recalls the history of our work on this topic, which resulted in the implementation of PRP and various variants of a saved leave scheme. In a future issue of Echo, we will come back on other elements of CERN's social policy that we would like to improve upon to ensure a better work-life balance.  Indeed, a good employer must offer a working environment that allows an...

  10. Beyond accreditation: excellence in medical education.

    Science.gov (United States)

    Ahn, Eusang; Ahn, Ducksun

    2014-01-01

    Medical school accreditation is a relatively new phenomenon in Korea. The development of an accreditation body and standards for a two-tiered "Must" and "Should" system in 1997 eventually led to the implementation of a third "Excellence" level of attainment. These standards were conceived out of a desire to be able to first recognize and promote outstanding performance of medical schools, second to provide role models in medical education, and furthermore to preview the third level as potential components of the pre-existing second level for the next accreditation cycle. It is a quality-assurance mechanism that, while not required for accreditation itself, pushes medical schools to go beyond the traditional requirements of mere pass-or-fail accreditation adequacy, and encourages schools to deliver an unprecedented level of medical education. The Association for Medical Education in Europe developed its own third-tier system of evaluation under the ASPIRE project, with many similar goals. Due to its advanced nature and global scope, the Korean accreditation body has decided to implement the ASPIRE system in Korea as well.

  11. Firm characteristics and excellence in integrated reporting

    Directory of Open Access Journals (Sweden)

    Natasha Buitendag

    2017-02-01

    Full Text Available Background: Integrated reporting has attracted much attention in the past few years, and South Africa has taken the lead in its development worldwide. An annual survey is published by Ernst & Young regarding the quality of the integrated reports of the top 100 entities listed on the Johannesburg Stock Exchange (JSE. Aim: The study on which this article is based was aimed at determining whether the assessment of an entity’s characteristics can predetermine the quality of the integrated report generated by that entity. Setting: This article focuses on an analysis of the integrated reporting of the top 100 entities listed on JSE for the financial years ending in 2013, 2014 and 2015. Methods: Comparison of categorical variables, mixed-model repeated measures ANOVA and generalised estimating equations were applied to identify the best classificators to distinguish between excellent integrated reporting and those reports where progress could still be made. Results: The results show that the type of industry the entity finds itself in, the size and profitability of the entity, as well as the composition of the members of the board, have an effect on the quality of the integrated report. Conclusion: Our results indicated that the type of industry, size of an entity, the profitability and composition of the board of directors, all have an effect on the quality of the integrated reporting. Our evidence will assist current and prospective stakeholders in evaluating the expected quality of an entity’s integrated report, through the evaluation of certain firm characteristics.

  12. 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.

  13. 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.)

  14. Pengembangan Modul Pembelajaran Pengolah Lembar Kerja Excel Berbasis Multimedia [Developing an Excel Spreadsheet Multimedia Learning Module

    Directory of Open Access Journals (Sweden)

    Yanuard Putro Dwikristanto

    2018-01-01

    Full Text Available Competence in the use of Information and Communication Technology (ICT has become a part that is needed to be mastered by prospective student-teachers. Students are expected to use ICT appropriately in education. The problem that often happens is that many students have difficulty working on tasks outside the classroom, especially when Microsoft Excel. Many students appear to understand and look like they are able follow guided practice in class, but outside of class they are confused and often forget the instructions given by lecturers. One solution for dealing with these problems is to provide learners with a multimedia module for learning how to create and use spreadsheets. It is hoped that this module will help students to solve their learning difficulties in ICT courses including those with different learning styles. The purpose of this research is to evaluate the usefulness of this multimedia module. The development of this module was done by using the ASSURE model which consists of six stages: (1 analyze the learners; (2 state objectives; (3 select appropriate methods, media, and materials; (4 utilize materials; (5 require learners’ participation; and (6 evaluate and revise. The results of the questionnaire indicate that candidate teachers tend to be stronger in learning through Visual-Auditory ways. In addition, student mastery of the Excel Sheet processing program has the lowest value of Word and PowerPoint programs. The design of this learning module will be made by taking into account the Visual-Auditory aspects of the Microsoft Excel topic. This learning module will be developed with an interactive design using the PowerPoint 2016 program.  BAHASA INDONESIA ABSTRAK: Kompetensi penggunaan Teknologi Informasi dan Komunikasi (TIK menjadi bagian yang perlu dikuasai oleh mahasiswa calon guru. Mahasiswa diharapkan dapat menggunakan TIK secara tepat dalam dunia pendidikan. Permasalahan yang sering terjadi adalah banyak mahasiswa yang

  15. Coding stimulus amplitude by correlated neural activity.

    Science.gov (United States)

    Metzen, Michael G; Ávila-Åkerberg, Oscar; Chacron, Maurice J

    2015-04-01

    While correlated activity is observed ubiquitously in the brain, its role in neural coding has remained controversial. Recent experimental results have demonstrated that correlated but not single-neuron activity can encode the detailed time course of the instantaneous amplitude (i.e., envelope) of a stimulus. These have furthermore demonstrated that such coding required and was optimal for a nonzero level of neural variability. However, a theoretical understanding of these results is still lacking. Here we provide a comprehensive theoretical framework explaining these experimental findings. Specifically, we use linear response theory to derive an expression relating the correlation coefficient to the instantaneous stimulus amplitude, which takes into account key single-neuron properties such as firing rate and variability as quantified by the coefficient of variation. The theoretical prediction was in excellent agreement with numerical simulations of various integrate-and-fire type neuron models for various parameter values. Further, we demonstrate a form of stochastic resonance as optimal coding of stimulus variance by correlated activity occurs for a nonzero value of noise intensity. Thus, our results provide a theoretical explanation of the phenomenon by which correlated but not single-neuron activity can code for stimulus amplitude and how key single-neuron properties such as firing rate and variability influence such coding. Correlation coding by correlated but not single-neuron activity is thus predicted to be a ubiquitous feature of sensory processing for neurons responding to weak input.

  16. Proposing an Environmental Excellence Self-Assessment Model

    DEFF Research Database (Denmark)

    Meulengracht Jensen, Peter; Johansen, John; Wæhrens, Brian Vejrum

    2013-01-01

    This paper presents an Environmental Excellence Self-Assessment (EEA) model based on the structure of the European Foundation of Quality Management Business Excellence Framework. Four theoretical scenarios for deploying the model are presented as well as managerial implications, suggesting...

  17. The Non-Destructive Determination of Burn-Up by Means of the Pr{sup l44} 2.18 M Gamma Activity

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R.S.; Blackadder, W.H.

    1965-05-15

    In recent years, gamma scanning has been used at several establishments for the determination of the burn-up profile along irradiated fuel elements, the 0.75 MeV gamma from Zr-95/Nb-95 being most often employed as the monitored radiation. Difficulties in establishing the geometry and the self-absorption of the gamma activity in the fuel have tended to prevent the application of the method to quantitative burn-up determination, which has usually been carried out by dissolution of selected portions of the fuel followed by conventional fission product separation or by uranium depletion methods. The present paper describes experiments carried out to calibrate a gamma scanner for quantitative measurements by counting the 2.18 MeV gamma activity due to Pr-144, the short-lived daughter of Ce-144 (t{sub 1/2} = 285 days) from selected pellets in several UO{sub 2} fuel specimens. Accurate burn-up values were then determined by dissolution and application of the isotopic dilution method, using stable molybdenum fission products. The elements, which were rotated about their longitudinal axes to minimize asymmetry effects, were viewed by a sodium iodide crystal and a multichannel analyser through a suitable collimator. Correction for attenuation of the gamma activity (much less than for 0.75 MeV) in the fuel elements which were of different diameters (12.6 to 15.04 mm) was made by applying relative attenuation factors and the effective geometry factor of the instrument was determined. In order to check the corrections applied, the counter factor was also calculated, for the 0.75 MeV activity from Zr-95/Nb-95 and in certain cases for the 0.66 MeV activity from Cs-137. The results obtained, demonstrate that at least over the range of diameters and cooling times used the method is suitable for quantitative determinations. Preliminary experiments to explore the possibility of using the high energy gammas (2.35, 2.65 MeV) from Rh-106 as a method for estimating the fraction of

  18. Driving performance excellence through disruptive innovation and visionary leadership

    OpenAIRE

    Coulson-Thomas, Colin

    2017-01-01

    Excellence and innovation are generally regarded as positive but they can mean different things to different people and both need to be relevant and affordable. This theme paper explores a variety of factors, issues and questions relating to excellence, innovation, leadership for excellence and innovation. It considers boardroom leadership and difficult choices, visionary boardroom leadership and factors to balance for responsible innovation, business excellence and the boardroom, driving exc...

  19. Technology cool women who code

    CERN Document Server

    Diehn, Andi

    2015-01-01

    Do you listen to music with an MP3 player or read books on a tablet? Do you play multiplayer video games with people on the other side of the world? Do you have a robot cleaning your kitchen? Maybe not yet, but someday! In Technology: Cool Women Who Code, kids in grades four through six learn about the thrilling effort that goes into researching, inventing, programming, and producing the technology we use today, from iPods to mechanical limbs. Young readers discover exactly what technology is, how it evolved, and where the future may lead. They also meet three women who have contributed to the field in critical ways, including Grace Hopper and Shaundra Bryant Daily. Technology: Cool Women Who Code combines high-interest content with links to online primary sources and essential questions that further expand kids' knowledge and understanding of a topic they come in contact with every day. Compelling portraits of women who have excelled in meeting the challenges of their field keep kids interested and infused w...

  20. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  1. QUALITY AWARDS: AN IMAGE OF BUSINESS EXCELLENCE

    Directory of Open Access Journals (Sweden)

    Ilies Liviu

    2015-07-01

    Full Text Available Across the world, increasingly more governmental organizations and industrial are doing everything possible to promote quality and to survive, the basic principle remains customer satisfaction and even more than that, it speaks of the principle of customer delight. In this sense, quality has become the source of sustained competitive advantage that provides organizations the supremacy of the global markets characterized by competition which becoming more and more intensified. Juran, one of the highest quality gurus say that “just as the twentieth century was the century of productivity, the twenty-first century will be the quality century” which is a very relevant and comprehensive statement of the economic reality of the past and a profound forecast for future business of the twenty-first century. In this regard, in order to achieve this competitive advantage, quality must be managed and this is accomplished through Total Quality Management (TQM. Quality awards models are instruments of total quality management through which quality can be assessed and improved, thus, knowing the quality awards models is critical for findings the new ways to improve the quality and performance of the organizations. The present paper aims to illustrate the best practices on quality improvement in this respect we intend to present the general framework of the quality awards for business excellence. In this sense we present the most important international quality awards, namely: "Malcolm Baldrige National Quality Award", "European Quality Award" and “Romanian Quality Award J. M. Juran". For this purpose we used as main sources of analyzing the structure and the operation mode of these three important quality awards, Juran's work (which is probably the most important work in the field of quality and other relevant sources in total quality management which treats issues related to quality awards and also we used as sources of updated information the official

  2. [Tabular excel editor for analysis of aligned nucleotide sequences].

    Science.gov (United States)

    Demkin, V V

    2010-01-01

    Excel platform was used for transition of results of multiple aligned nucleotide sequences obtained using the BLAST network service to the form appropriate for visual analysis and editing. Two macros operators for MS Excel 2007 were constructed. The array of aligned sequences transformed into Excel table and processed using macros operators is more appropriate for analysis than initial html data.

  3. In Search of Excellence: Historical Roots of Greek Culture.

    Science.gov (United States)

    Makedon, Alexander

    This paper seeks to discover conditions that motivate people to achieve excellence and uses the Greek culture as an example of excellence. The document addresses the basic questions: (1) What were the social conditions that resulted in what is widely known as the "Greek Miracle"?; (2) What motivated the ancient Greeks to excel,…

  4. Overcoming Microsoft Excel's Weaknesses for Crop Model Building and Simulations

    Science.gov (United States)

    Sung, Christopher Teh Boon

    2011-01-01

    Using spreadsheets such as Microsoft Excel for building crop models and running simulations can be beneficial. Excel is easy to use, powerful, and versatile, and it requires the least proficiency in computer programming compared to other programming platforms. Excel, however, has several weaknesses: it does not directly support loops for iterative…

  5. 77 FR 33203 - Equity and Excellence Commission, Meeting Cancellation

    Science.gov (United States)

    2012-06-05

    ... Equity and Excellence Commission, Meeting Cancellation AGENCY: U.S. Department of Education. ACTION... cancellation of the Meeting of the Equity and Excellence Commission scheduled for June 4, 2012 and announced in..., Equity and Excellence Commission, U.S. Department of Education, 400 Maryland Avenue SW., Washington, DC...

  6. Excellence: The Importance of Vision and Work Ethic

    Science.gov (United States)

    Hill, Roger B.

    2010-01-01

    In this paper, the author discusses the topic of excellence and shares key aspects of excellence that he has found to be true in his life's journey. He discusses two elements of excellence, namely (1) "vision"; and (2) "work ethic", and describes some characteristics of each element. Three aspects: (1) interpersonal skills; (2) initiative; and (3)…

  7. Attributes for Measuring Equity and Excellence in District Operation.

    Science.gov (United States)

    DeMoulin, Donald F.; Guyton, John W.

    In the quest for excellence, school districts have a variety of indicators or attributes available by which to gauge their progress. This model, used By the Equity and Excellence Research school districts in Mississippi, monitors achievement in relation to educational excellence. Team members established a list of attributes and various means of…

  8. Essentials of Excel, Excel VBA, SAS and Minitab for statistical and financial analyses

    CERN Document Server

    Lee, Cheng-Few; Chang, Jow-Ran; Tai, Tzu

    2016-01-01

    This introductory textbook for business statistics teaches statistical analysis and research methods via business case studies and financial data using Excel, MINITAB, and SAS. Every chapter in this textbook engages the reader with data of individual stock, stock indices, options, and futures. One studies and uses statistics to learn how to study, analyze, and understand a data set of particular interest. Some of the more popular statistical programs that have been developed to use statistical and computational methods to analyze data sets are SAS, SPSS, and MINITAB. Of those, we look at MINITAB and SAS in this textbook. One of the main reasons to use MINITAB is that it is the easiest to use among the popular statistical programs. We look at SAS because it is the leading statistical package used in industry. We also utilize the much less costly and ubiquitous Microsoft Excel to do statistical analysis, as the benefits of Excel have become widely recognized in the academic world and its analytical capabilities...

  9. The Aesthetics of Coding

    DEFF Research Database (Denmark)

    Andersen, Christian Ulrik

    2007-01-01

    Computer art is often associated with computer-generated expressions (digitally manipulated audio/images in music, video, stage design, media facades, etc.). In recent computer art, however, the code-text itself – not the generated output – has become the artwork (Perl Poetry, ASCII Art, obfuscated...... code, etc.). The presentation relates this artistic fascination of code to a media critique expressed by Florian Cramer, claiming that the graphical interface represents a media separation (of text/code and image) causing alienation to the computer’s materiality. Cramer is thus the voice of a new ‘code...... avant-garde’. In line with Cramer, the artists Alex McLean and Adrian Ward (aka Slub) declare: “art-oriented programming needs to acknowledge the conditions of its own making – its poesis.” By analysing the Live Coding performances of Slub (where they program computer music live), the presentation...

  10. Opening up codings?

    DEFF Research Database (Denmark)

    Steensig, Jakob; Heinemann, Trine

    2015-01-01

    doing formal coding and when doing more “traditional” conversation analysis research based on collections. We are more wary, however, of the implication that coding-based research is the end result of a process that starts with qualitative investigations and ends with categories that can be coded......We welcome Tanya Stivers’s discussion (Stivers, 2015/this issue) of coding social interaction and find that her descriptions of the processes of coding open up important avenues for discussion, among other things of the precise ad hoc considerations that researchers need to bear in mind, both when....... Instead we propose that the promise of coding-based research lies in its ability to open up new qualitative questions....

  11. Overview of Code Verification

    Science.gov (United States)

    1983-01-01

    The verified code for the SIFT Executive is not the code that executes on the SIFT system as delivered. The running versions of the SIFT Executive contain optimizations and special code relating to the messy interface to the hardware broadcast interface and to packing of data to conserve space in the store of the BDX930 processors. The running code was in fact developed prior to and without consideration of any mechanical verification. This was regarded as necessary experimentation with the SIFT hardware and special purpose Pascal compiler. The Pascal code sections cover: the selection of a schedule from the global executive broadcast, scheduling, dispatching, three way voting, and error reporting actions of the SIFT Executive. Not included in these sections of Pascal code are: the global executive, five way voting, clock synchronization, interactive consistency, low level broadcasting, and program loading, initialization, and schedule construction.

  12. Phonological coding during reading

    Science.gov (United States)

    Leinenger, Mallorie

    2014-01-01

    The exact role that phonological coding (the recoding of written, orthographic information into a sound based code) plays during silent reading has been extensively studied for more than a century. Despite the large body of research surrounding the topic, varying theories as to the time course and function of this recoding still exist. The present review synthesizes this body of research, addressing the topics of time course and function in tandem. The varying theories surrounding the function of phonological coding (e.g., that phonological codes aid lexical access, that phonological codes aid comprehension and bolster short-term memory, or that phonological codes are largely epiphenomenal in skilled readers) are first outlined, and the time courses that each maps onto (e.g., that phonological codes come online early (pre-lexical) or that phonological codes come online late (post-lexical)) are discussed. Next the research relevant to each of these proposed functions is reviewed, discussing the varying methodologies that have been used to investigate phonological coding (e.g., response time methods, reading while eyetracking or recording EEG and MEG, concurrent articulation) and highlighting the advantages and limitations of each with respect to the study of phonological coding. In response to the view that phonological coding is largely epiphenomenal in skilled readers, research on the use of phonological codes in prelingually, profoundly deaf readers is reviewed. Finally, implications for current models of word identification (activation-verification model (Van Order, 1987), dual-route model (e.g., Coltheart, Rastle, Perry, Langdon, & Ziegler, 2001), parallel distributed processing model (Seidenberg & McClelland, 1989)) are discussed. PMID:25150679

  13. The aeroelastic code FLEXLAST

    Energy Technology Data Exchange (ETDEWEB)

    Visser, B. [Stork Product Eng., Amsterdam (Netherlands)

    1996-09-01

    To support the discussion on aeroelastic codes, a description of the code FLEXLAST was given and experiences within benchmarks and measurement programmes were summarized. The code FLEXLAST has been developed since 1982 at Stork Product Engineering (SPE). Since 1992 FLEXLAST has been used by Dutch industries for wind turbine and rotor design. Based on the comparison with measurements, it can be concluded that the main shortcomings of wind turbine modelling lie in the field of aerodynamics, wind field and wake modelling. (au)

  14. Generating code adapted for interlinking legacy scalar code and extended vector code

    Science.gov (United States)

    Gschwind, Michael K

    2013-06-04

    Mechanisms for intermixing code are provided. Source code is received for compilation using an extended Application Binary Interface (ABI) that extends a legacy ABI and uses a different register configuration than the legacy ABI. First compiled code is generated based on the source code, the first compiled code comprising code for accommodating the difference in register configurations used by the extended ABI and the legacy ABI. The first compiled code and second compiled code are intermixed to generate intermixed code, the second compiled code being compiled code that uses the legacy ABI. The intermixed code comprises at least one call instruction that is one of a call from the first compiled code to the second compiled code or a call from the second compiled code to the first compiled code. The code for accommodating the difference in register configurations is associated with the at least one call instruction.

  15. Decoding of Cyclic Codes,

    Science.gov (United States)

    INFORMATION THEORY, *DECODING), (* DATA TRANSMISSION SYSTEMS , DECODING), STATISTICAL ANALYSIS, STOCHASTIC PROCESSES, CODING, WHITE NOISE, NUMBER THEORY, CORRECTIONS, BINARY ARITHMETIC, SHIFT REGISTERS, CONTROL SYSTEMS, USSR

  16. ARC Code TI: ACCEPT

    Data.gov (United States)

    National Aeronautics and Space Administration — ACCEPT consists of an overall software infrastructure framework and two main software components. The software infrastructure framework consists of code written to...

  17. Diameter Perfect Lee Codes

    CERN Document Server

    Horak, Peter

    2011-01-01

    Lee codes have been intensively studied for more than 40 years. Interest in these codes has been triggered by the Golomb-Welch conjecture on the existence of perfect error-correcting Lee codes. In this paper we deal with the existence and enumeration of diameter perfect Lee codes. As main results we determine all q for which there exists a linear diameter-4 perfect Lee code of word length n over Z_{q}, and prove that for each n\\geq3 there are unaccountably many diameter-4 perfect Lee codes of word length n over Z. This is in a strict contrast with perfect error-correcting Lee codes of word length n over Z as there is a unique such code for n=3, and its is conjectured that this is always the case when 2n+1 is a prime. Diameter perfect Lee codes will be constructed by an algebraic construction that is based on a group homomorphism. This will allow us to design an efficient algorithm for their decoding.

  18. Expander chunked codes

    Science.gov (United States)

    Tang, Bin; Yang, Shenghao; Ye, Baoliu; Yin, Yitong; Lu, Sanglu

    2015-12-01

    Chunked codes are efficient random linear network coding (RLNC) schemes with low computational cost, where the input packets are encoded into small chunks (i.e., subsets of the coded packets). During the network transmission, RLNC is performed within each chunk. In this paper, we first introduce a simple transfer matrix model to characterize the transmission of chunks and derive some basic properties of the model to facilitate the performance analysis. We then focus on the design of overlapped chunked codes, a class of chunked codes whose chunks are non-disjoint subsets of input packets, which are of special interest since they can be encoded with negligible computational cost and in a causal fashion. We propose expander chunked (EC) codes, the first class of overlapped chunked codes that have an analyzable performance, where the construction of the chunks makes use of regular graphs. Numerical and simulation results show that in some practical settings, EC codes can achieve rates within 91 to 97 % of the optimum and outperform the state-of-the-art overlapped chunked codes significantly.

  19. QR codes for dummies

    CERN Document Server

    Waters, Joe

    2012-01-01

    Find out how to effectively create, use, and track QR codes QR (Quick Response) codes are popping up everywhere, and businesses are reaping the rewards. Get in on the action with the no-nonsense advice in this streamlined, portable guide. You'll find out how to get started, plan your strategy, and actually create the codes. Then you'll learn to link codes to mobile-friendly content, track your results, and develop ways to give your customers value that will keep them coming back. It's all presented in the straightforward style you've come to know and love, with a dash of humor thrown

  20. Academic excellence workshops in chemistry and physics

    Science.gov (United States)

    Mills, Susan Rose

    In the mid-1970's, Uri Treisman, at the University of California, Berkeley, developed an academic excellence workshop program that had important successes in increasing minority student achievement and persistence in calculus. The present dissertation research is an in-depth study of chemistry and physics workshops at the California State Polytechnic University, Pomona. Data for the first, longitudinal component of this study were obtained by tracking to Spring 1998 all workshop minority students, i.e., Latino, African American, and Native American workshop students, a random sample of non-workshop minority students, and a random sample of non-targeted students, i.e., Anglo and Asian students, enrolled in first-quarter General Chemistry or Physics during specific quarters of 1992 or 1993. Data for the second component were obtained by administering questionnaires, conducting interviews, and observing science students during Fall, 1996. Workshop participation was a significant predictor of first-quarter course grade for minority students in both chemistry and physics, while verbal and mathematics Scholastic Aptitude Test (SAT) scores were not significant predictors of beginning course grade for minority science students. The lack of predictive ability of the SAT and the importance of workshop participation in minority students' beginning science course performance are results with important implications for educators and students. In comparing pre-college achievement measures for workshop and non-targeted students, non-targeted students' mathematics SAT scores were significantly higher than chemistry and physics workshop students' scores. Nonetheless, workshop participation "leveled the field" as workshop and non-targeted students performed similarly in beginning science courses. Positive impacts of workshop participation on achievement, persistence, efficiency, social integration, and self-confidence support the continued and expanded funding of workshop programs

  1. Excel 2013 all-in-one for dummies

    CERN Document Server

    Harvey, Greg

    2013-01-01

    The comprehensive reference, now completely up-to-date for Excel 2013! As the standard for spreadsheet applications, Excel is used worldwide - but it's not always user-friendly. However, in the hands of veteran bestselling author Greg Harvey, Excel gets a whole lot easier to understand! This handy all-in-one guide covers all the essentials, the new features, how to analyze data with Excel, and much more. The featured minibooks address Excel basics, worksheet design, formulas and functions, worksheet collaboration and review, charts and graphics, data management, data analysis, and

  2. Development and preliminary verification of the 3D core neutronic code: COCO

    Energy Technology Data Exchange (ETDEWEB)

    Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J. [Reactor Design and Fuel Management Research Center, China Nuclear Power Technology Research Inst., 47F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen (China)

    2012-07-01

    As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

  3. On {\\sigma}-LCD codes

    OpenAIRE

    Carlet, Claude; Mesnager, Sihem; Tang, Chunming; Qi, Yanfeng

    2017-01-01

    Linear complementary pairs (LCP) of codes play an important role in armoring implementations against side-channel attacks and fault injection attacks. One of the most common ways to construct LCP of codes is to use Euclidean linear complementary dual (LCD) codes. In this paper, we first introduce the concept of linear codes with $\\sigma$ complementary dual ($\\sigma$-LCD), which includes known Euclidean LCD codes, Hermitian LCD codes, and Galois LCD codes. As Euclidean LCD codes, $\\sigma$-LCD ...

  4. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  5. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  6. SNP_tools: A compact tool package for analysis and conversion of genotype data for MS-Excel.

    Science.gov (United States)

    Chen, Bowang; Wilkening, Stefan; Drechsel, Marion; Hemminki, Kari

    2009-10-23

    Single nucleotide polymorphism (SNP) genotyping is a major activity in biomedical research. Scientists prefer to have a facile access to the results which may require conversions between data formats. First hand SNP data is often entered in or saved in the MS-Excel format, but this software lacks genetic and epidemiological related functions. A general tool to do basic genetic and epidemiological analysis and data conversion for MS-Excel is needed. The SNP_tools package is prepared as an add-in for MS-Excel. The code is written in Visual Basic for Application, embedded in the Microsoft Office package. This add-in is an easy to use tool for users with basic computer knowledge (and requirements for basic statistical analysis). Our implementation for Microsoft Excel 2000-2007 in Microsoft Windows 2000, XP, Vista and Windows 7 beta can handle files in different formats and converts them into other formats. It is a free software.

  7. Scrum Code Camps

    DEFF Research Database (Denmark)

    Pries-Heje, Jan; Pries-Heje, Lene; Dahlgaard, Bente

    2013-01-01

    is required. In this paper we present the design of such a new approach, the Scrum Code Camp, which can be used to assess agile team capability in a transparent and consistent way. A design science research approach is used to analyze properties of two instances of the Scrum Code Camp where seven agile teams...

  8. Error Correcting Codes

    Indian Academy of Sciences (India)

    be fixed to define codes over such domains). New decoding schemes that take advantage of such connections can be devised. These may soon show up in a technique called code division multiple access (CDMA) which is proposed as a basis for digital cellular communication. CDMA provides a facility for many users to ...

  9. Codes of Conduct

    Science.gov (United States)

    Million, June

    2004-01-01

    Most schools have a code of conduct, pledge, or behavioral standards, set by the district or school board with the school community. In this article, the author features some schools that created a new vision of instilling code of conducts to students based on work quality, respect, safety and courtesy. She suggests that communicating the code…

  10. Error Correcting Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 2; Issue 3. Error Correcting Codes - Reed Solomon Codes. Priti Shankar. Series Article Volume 2 Issue 3 March 1997 pp 33-47. Fulltext. Click here to view fulltext PDF. Permanent link: http://www.ias.ac.in/article/fulltext/reso/002/03/0033-0047 ...

  11. Code Generation = A* + BURS

    NARCIS (Netherlands)

    Nymeyer, Albert; Katoen, Joost P.; Westra, Ymte; Alblas, H.; Gyimóthy, Tibor

    1996-01-01

    A system called BURS that is based on term rewrite systems and a search algorithm A* are combined to produce a code generator that generates optimal code. The theory underlying BURS is re-developed, formalised and explained in this work. The search algorithm uses a cost heuristic that is derived

  12. Dress Codes for Teachers?

    Science.gov (United States)

    Million, June

    2004-01-01

    In this article, the author discusses an e-mail survey of principals from across the country regarding whether or not their school had a formal staff dress code. The results indicate that most did not have a formal dress code, but agreed that professional dress for teachers was not only necessary, but showed respect for the school and had a…

  13. Informal control code logic

    NARCIS (Netherlands)

    Bergstra, J.A.

    2010-01-01

    General definitions as well as rules of reasoning regarding control code production, distribution, deployment, and usage are described. The role of testing, trust, confidence and risk analysis is considered. A rationale for control code testing is sought and found for the case of safety critical

  14. Interleaved Product LDPC Codes

    OpenAIRE

    Baldi, Marco; Cancellieri, Giovanni; Chiaraluce, Franco

    2011-01-01

    Product LDPC codes take advantage of LDPC decoding algorithms and the high minimum distance of product codes. We propose to add suitable interleavers to improve the waterfall performance of LDPC decoding. Interleaving also reduces the number of low weight codewords, that gives a further advantage in the error floor region.

  15. Nuremberg code turns 60

    OpenAIRE

    Thieren, Michel; Mauron, Alexandre

    2007-01-01

    This month marks sixty years since the Nuremberg code – the basic text of modern medical ethics – was issued. The principles in this code were articulated in the context of the Nuremberg trials in 1947. We would like to use this anniversary to examine its ability to address the ethical challenges of our time.

  16. Error Correcting Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 2; Issue 1. Error Correcting Codes The Hamming Codes. Priti Shankar. Series Article Volume 2 Issue 1 January ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science, Bangalore 560 012, India ...

  17. Experimental results and analysis of core physics experiments, FUBILA, for high burn-up BWR full MOX cores

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, T.; Kikuchi, S.; Kawashima, K.; Kamimura, K. [Japan Nuclear Energy Safety Organization, 3-17-1, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2006-07-01

    JNES has been performing MOX core physics experiments, FUBILA, in the EOLE critical facility of the CEA Cadarache center with collaboration of a French Consortium (CEA and COGEMA). The experiments have been designed to obtain the core physics data of operating conditions of full MOX BWR cores consisting of high burn up BWR MOX assemblies. The experiments consisting of seven different core configurations started from January 2005 and will be completed by August 2006. Theoretical analysis of the experimental data has been also carried out using a deterministic code, SRAC, and a continuous energy Monte Carlo calculation code, MVP, with major nuclear data libraries, JENDL-3.3, 3.2, ENDF/B-VI and JEFF-3.1 for the first critical core. (authors)

  18. Impact of neutron thermal scattering laws on the burn-up analysis of supercritical LWR's fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Conti, Andrea

    2011-10-15

    . The third, most naive option is called the ''free gas approximation''. It is the goal of this work to make an estimate of the criticality calculations' inaccuracy due to the inadequate employed physical model and to determine which one of the available models can be the best replacement. The accuracy of criticality calculations referring to the HPLWR is a problem that had already been raised by Waata in 2006. In her Ph.D. thesis Waata reports having carried out MCNP runs referring to an HPLWR fuel element employing the free gas approximation. In her thesis Waata explicitly sifts through the factors that can affect her MCNP runs' accuracy, but leaves the inappropriate thermal treatment completely out. In this work, the inaccuracy of the criticality calculations has been investigated carrying out sets of similar burn-up calculations differing from each other only in the applied thermal cross section sets. The widest discrepancies were detected between the results obtained applying the free gas model and those obtained applying the molecular models. This, in conjunction with the fact that the free gas model does not even keep in count the molecular structure of H{sub 2}O suggest to discard it and to focus the investigation on the vapour and liquid models. Dr. J. Marti, from the Universitat Politecnica de Catalunya, Barcelona, Spain registered the generalized frequency distributions obtained from the molecular dynamics simulations of 216 molecules of H{sub 2}O in 10 simulated supercritical states and published in an article (1999) the frequencies of the three characteristic distribution peaks for each simulated state, in numerical format. A confrontation with the corresponding peaks from Bernnat's available frequency distributions for liquid water and vapour revealed the peaks of the latter to be closest to the supercritical water ones in nearly all cases. Hence the inference that thermal cross section sets for vapour are for the time

  19. Evaluation Codes from an Affine Veriety Code Perspective

    DEFF Research Database (Denmark)

    Geil, Hans Olav

    2008-01-01

    Evaluation codes (also called order domain codes) are traditionally introduced as generalized one-point geometric Goppa codes. In the present paper we will give a new point of view on evaluation codes by introducing them instead as particular nice examples of affine variety codes. Our study...... . . . . . . . . . . . . . . . . . . . . . . . 171 4.9 Codes form order domains . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 4.10 One-point geometric Goppa codes . . . . . . . . . . . . . . . . . . . . . . . . 176 4.11 Bibliographical Notes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 178 References...

  20. Quantum Synchronizable Codes From Quadratic Residue Codes and Their Supercodes

    OpenAIRE

    Xie, Yixuan; Yuan, Jinhong; Fujiwara, Yuichiro

    2014-01-01

    Quantum synchronizable codes are quantum error-correcting codes designed to correct the effects of both quantum noise and block synchronization errors. While it is known that quantum synchronizable codes can be constructed from cyclic codes that satisfy special properties, only a few classes of cyclic codes have been proved to give promising quantum synchronizable codes. In this paper, using quadratic residue codes and their supercodes, we give a simple construction for quantum synchronizable...

  1. Clinical excellence in nephrology: Examples from the published literature.

    Science.gov (United States)

    Geetha, Duvuru; Lee, Steven K; Srivastava, Amar J; Kraus, Edward S; Wright, Scott M

    2015-08-15

    Provision of exceptional medical care is a goal for the medical profession because this is what the public needs and deserves. Academic medical centers that value excellent clinicians may have the best chance to recruit and retain these faculty members. When our institution hoped to launch the Miller Coulson Academy of Clinical Excellence to measure and reward master clinicians, a critical first step was to use rigorous methods to develop a definition of clinical excellence. Published papers have illustrated that this general definition of clinical excellence is applicable to fields of psychiatry, cardiology, and pediatrics. In this manuscript, we apply the definition of clinical excellence to nephrology. Using the same framework, we reviewed the literature to find clinical cases and exemplary nephrologists that highlight the specific domains. This collection of reports in nephrology illustrates that the definition of clinical excellence set forth by the Miller Coulson Academy is highly applicable to physicians caring for individuals with kidney disease. Relating the definition of clinical excellence to renal medicine is worthwhile in that it can help to exemplify the model to which physicians and trainees may seek to aspire. Many examples of clinical excellence in renal medicine can be found in the published medical literature. The domains of clinical excellence, described by the Miller-Coulson Academy of Clinical Excellence, apply very well to the field of nephrology.

  2. Modelización financiera con Excel

    Directory of Open Access Journals (Sweden)

    Lafuente Robledo, M.

    1998-01-01

    puramente didáctico facilita la comprensión y el estudio del alumno, el cual se tiene que centrar en los aspectos básicos de la asignatura, reduciéndose de forma muy importante la memorización de conceptos. · Permite determinar de manera rápida y exacta los distintos valores que sea necesario obtener y, consecuentemente, comparar entre distintas operaciones financieras planteadas al alumno de forma inmediata. · Operaciones con cálculos numéricos importantes, como la realización de cuadros de amortización de préstamos y empréstitos, también se pueden realizar con gran rapidez, y llevar a cabo un análisis de sensibilidad del mismo. · En general, permite resolver el problema hasta la solución final y extraer las conclusiones operativas, fase que en los cálculos manuales es muchas veces soslayada. Consecuentemente, facilita la detección de posibles inconsistencias y datos erróneos en los planteamientos y resultados. · La utilización de herramientas informáticas ayuda por otra parte a motivar a los alumnos como consecuencia del atractivo generalizado del soporte informático. No obstante, este aspecto no se puede generalizar a todos los alumnos, y además tiene un efecto que se diluye con el tiempo. A continuación, se desarrollan una serie de modelos de matemática financiera con hoja de cálculo con el objeto de ilustrar de forma clara y precisa como se pueden implementar dichos problemas financieros en Excel. Los tres primeros casos corresponden a la matemática financiera clásica y consideramos que deberían representar las aplicaciones informáticas mínimas de cualquier curso de esta disciplina en la Facultades y Escuelas Universitarias de Ciencias Empresariales.

  3. [MapDraw: a microsoft excel macro for drawing genetic linkage maps based on given genetic linkage data].

    Science.gov (United States)

    Liu, Ren-Hu; Meng, Jin-Ling

    2003-05-01

    MAPMAKER is one of the most widely used computer software package for constructing genetic linkage maps.However, the PC version, MAPMAKER 3.0 for PC, could not draw the genetic linkage maps that its Macintosh version, MAPMAKER 3.0 for Macintosh,was able to do. Especially in recent years, Macintosh computer is much less popular than PC. Most of the geneticists use PC to analyze their genetic linkage data. So a new computer software to draw the same genetic linkage maps on PC as the MAPMAKER for Macintosh to do on Macintosh has been crying for. Microsoft Excel,one component of Microsoft Office package, is one of the most popular software in laboratory data processing. Microsoft Visual Basic for Applications (VBA) is one of the most powerful functions of Microsoft Excel. Using this program language, we can take creative control of Excel, including genetic linkage map construction, automatic data processing and more. In this paper, a Microsoft Excel macro called MapDraw is constructed to draw genetic linkage maps on PC computer based on given genetic linkage data. Use this software,you can freely construct beautiful genetic linkage map in Excel and freely edit and copy it to Word or other application. This software is just an Excel format file. You can freely copy it from ftp://211.69.140.177 or ftp://brassica.hzau.edu.cn and the source code can be found in Excel's Visual Basic Editor.

  4. Neutron spallation source and the Dubna cascade code

    CERN Document Server

    Kumar, V; Goel, U; Barashenkov, V S

    2003-01-01

    Neutron multiplicity per incident proton, n/p, in collision of high energy proton beam with voluminous Pb and W targets has been estimated from the Dubna cascade code and compared with the available experimental data for the purpose of benchmarking of the code. Contributions of various atomic and nuclear processes for heat production and isotopic yield of secondary nuclei are also estimated to assess the heat and radioactivity conditions of the targets. Results obtained from the code show excellent agreement with the experimental data at beam energy, E < 1.2 GeV and differ maximum up to 25% at higher energy. (author)

  5. Pyramid image codes

    Science.gov (United States)

    Watson, Andrew B.

    1990-01-01

    All vision systems, both human and machine, transform the spatial image into a coded representation. Particular codes may be optimized for efficiency or to extract useful image features. Researchers explored image codes based on primary visual cortex in man and other primates. Understanding these codes will advance the art in image coding, autonomous vision, and computational human factors. In cortex, imagery is coded by features that vary in size, orientation, and position. Researchers have devised a mathematical model of this transformation, called the Hexagonal oriented Orthogonal quadrature Pyramid (HOP). In a pyramid code, features are segregated by size into layers, with fewer features in the layers devoted to large features. Pyramid schemes provide scale invariance, and are useful for coarse-to-fine searching and for progressive transmission of images. The HOP Pyramid is novel in three respects: (1) it uses a hexagonal pixel lattice, (2) it uses oriented features, and (3) it accurately models most of the prominent aspects of primary visual cortex. The transform uses seven basic features (kernels), which may be regarded as three oriented edges, three oriented bars, and one non-oriented blob. Application of these kernels to non-overlapping seven-pixel neighborhoods yields six oriented, high-pass pyramid layers, and one low-pass (blob) layer.

  6. Report number codes

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.N. (ed.)

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name.

  7. Utforming av innendeltaker-design grafer i Microsoft Excel 2007

    OpenAIRE

    Strømgren, Børge; Løkke, Gunn; Olsen, Fie; Løkke, Jon Arne

    2010-01-01

    I artikkelen gis det beskrivelser av hvordan grafer for framstilling av innendeltaker tilba-ketrekkingsingsdesign (reversering), multielementdesign og multippel basislinjedesign kan konstrueres i Microsoft Excel 2007 for Windows. Det gis tips om hvordan ferdige grafer kan overføres fra Excel til andre relevante programmer. Forutsetning for bruk av framstillingen er korrekt installert Microsoft Excel 2007 for Windows. Beskrivelsene er tenkt å gi utfyllende informasjon om konstruksjon av innend...

  8. Measurement of excellence and cost in a supply chain

    OpenAIRE

    Pettersson, Annelie

    2015-01-01

    This Doctoral Thesis addresses the topic of measurement of excellence and cost in a supply chain. In this thesis, a framework for monitoring and analysis of cost in the supply chain is developed. A model for analyzing excellence in a supply chain is also developed. The work aims to facilitate improvements and to improve decisionmaking regarding Supply Chain Management (SCM) and excellence and cost improvements in supply chains. The thesis contains an introduction, a literature review and five...

  9. Cryptography cracking codes

    CERN Document Server

    2014-01-01

    While cracking a code might seem like something few of us would encounter in our daily lives, it is actually far more prevalent than we may realize. Anyone who has had personal information taken because of a hacked email account can understand the need for cryptography and the importance of encryption-essentially the need to code information to keep it safe. This detailed volume examines the logic and science behind various ciphers, their real world uses, how codes can be broken, and the use of technology in this oft-overlooked field.

  10. Quantum coding theorems

    Science.gov (United States)

    Holevo, A. S.

    1998-12-01

    ContentsI. IntroductionII. General considerations § 1. Quantum communication channel § 2. Entropy bound and channel capacity § 3. Formulation of the quantum coding theorem. Weak conversionIII. Proof of the direct statement of the coding theorem § 1. Channels with pure signal states § 2. Reliability function § 3. Quantum binary channel § 4. Case of arbitrary states with bounded entropyIV. c-q channels with input constraints § 1. Coding theorem § 2. Gauss channel with one degree of freedom § 3. Classical signal on quantum background noise Bibliography

  11. Excel 2003 all-in-one desk reference for dummies

    CERN Document Server

    Harvey, Greg

    2013-01-01

    When you think of number-crunching and spreadsheets, you think of Excel, right? After Word, it's the most popular program in the Microsoft Office suite. But if technical jargon isn't your first language, you may have found Excel just a teeny bit frustrating. It can be really hard to pick your way through the many features and make Excel do what you need for it to do. Once you know how, you can use Excel to Create fill-in-the-blank forms Prepare expense reports and invoices Manage all sorts of data Keep sales and inventory records Analyze financial data and create forecasts Present informati

  12. Teaching excellence in nursing education: a caring framework.

    Science.gov (United States)

    Sawatzky, Jo-Ann V; Enns, Carol L; Ashcroft, Terri J; Davis, Penny L; Harder, B Nicole

    2009-01-01

    Nursing education plays a central role in the ability to practice effectively. It follows that an optimally educated nursing workforce begets optimal patient care. A framework for excellence in nursing education could guide the development of novice educators, establish the basis for evaluating teaching excellence, and provide the impetus for research in this area. However, a review of the social sciences and nursing literature as well as a search for existing models for teaching excellence revealed an apparent dearth of evidence specific to excellence in nursing education. Therefore, we developed the Caring Framework for Excellence in Nursing Education. This framework evolved from a review of the generic constructs that exemplify teaching excellence: excellence in teaching practice, teaching scholarship, and teaching leadership. Nursing is grounded in the ethic of caring. Hence, caring establishes the foundation for this uniquely nursing framework. Because a teaching philosophy is intimately intertwined with one's nursing philosophy and the ethic of caring, it is also fundamental to the caring framework. Ideally, this framework will contribute to excellence in nursing education and as a consequence excellence in nursing practice and optimal patient care.

  13. Head First Excel A learner's guide to spreadsheets

    CERN Document Server

    Milton, Michael

    2010-01-01

    Do you use Excel for simple lists, but get confused and frustrated when it comes to actually doing something useful with all that data? Stop tearing your hair out: Head First Excel helps you painlessly move from spreadsheet dabbler to savvy user. Whether you're completely new to Excel or an experienced user looking to make the program work better for you, this book will help you incorporate Excel into every aspect of your workflow, from a scratch pad for data-based brainstorming to exploratory analysis with PivotTables, optimizing outcomes with Goal Seek, and presenting your conclusions wit

  14. Brewer, Maine Wastewater Treatment Plant Recognized for Excellence

    Science.gov (United States)

    The Brewer Water Pollution Control Facility was recently honored with a 2015 Regional Wastewater Treatment Plant Excellence Award by the US Environmental Protection Agency's New England regional office.

  15. Design of a weapons-grade plutonium assembly for optimal burnup in a standard pressurized water reactor

    Science.gov (United States)

    Alonso-Vargas, Gustavo

    We created a new MOX fuel assembly design that can be used in standard Westinghouse pressurized water reactors (PWR) to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which is called MIX-33, appears to be a good option for the disposition of weapons-grade plutonium (WG-Pu), increasing the plutonium disposition rate by 8% compared to a previous Westinghouse design. It is based in two novel ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the increase of the moderation ratio of the assembly. We replaced 8 fuel pins by water holes to increase the moderation ratio. We can transition smoothly from a full LEU core to a full MIX-33 core meeting the operational and safety regulations of a standard PWR. Given a MOX supply interruption scenario we can transition smoothly to full LEU meeting the safety regulations and using standard LEU assemblies with uniform enriched pin-wise distribution. If the MOX supply is interrupted for only one cycle, we are able to transition back to full MIX-33 core. However, in this case we probably need to de-rate the power by a few percent for a few weeks at the beginning of the cycle (BOC) to accommodate high peaking. For comparison we created another assembly design without extra water holes, which we called "MIX-25". It behaves in all the conditions analyzed in a similar way to the MIX-33 but it does present minor control problems. These can be solved by making small modifications to the control and safety systems, namely by enriching the boron-10 content of some boron absorbers. Thus, the addition of water holes replacing fuel pins helps to improve the MIX-33 performance and eliminate the difficulties seen in the MIX-25 design. We also performed a benchmarking analysis to test the code CASMO-3 to analyze WG-Pu assemblies, using the code MCNP-4A to compare. We found good agreement between CASMO-3 and

  16. An Excel Macro to Plot the HFE-Diagram to Identify Sea Water Intrusion Phases.

    Science.gov (United States)

    Giménez-Forcada, Elena; Sánchez San Román, F Javier

    2015-01-01

    A hydrochemical facies evolution diagram (HFE-D) is a multirectangular diagram, which is a useful tool in the interpretation of sea water intrusion processes. This method note describes a simple method for generating an HFE-D plot using the spreadsheet software package, Microsoft Excel. The code was applied to groundwater from the alluvial coastal plain of Grosseto (Tuscany, Italy), which is characterized by a complex salinization process in which sea water mixes with sulfate or bicarbonate recharge water. © 2014, National GroundWater Association.

  17. Fulcrum Network Codes

    DEFF Research Database (Denmark)

    2015-01-01

    Fulcrum network codes, which are a network coding framework, achieve three objectives: (i) to reduce the overhead per coded packet to almost 1 bit per source packet; (ii) to operate the network using only low field size operations at intermediate nodes, dramatically reducing complexity...... in the network; and (iii) to deliver an end-to-end performance that is close to that of a high field size network coding system for high-end receivers while simultaneously catering to low-end ones that can only decode in a lower field size. Sources may encode using a high field size expansion to increase...... the number of dimensions seen by the network using a linear mapping. Receivers can tradeoff computational effort with network delay, decoding in the high field size, the low field size, or a combination thereof....

  18. VT ZIP Code Areas

    Data.gov (United States)

    Vermont Center for Geographic Information — (Link to Metadata) A ZIP Code Tabulation Area (ZCTA) is a statistical geographic entity that approximates the delivery area for a U.S. Postal Service five-digit...

  19. Bandwidth efficient coding

    CERN Document Server

    Anderson, John B

    2017-01-01

    Bandwidth Efficient Coding addresses the major challenge in communication engineering today: how to communicate more bits of information in the same radio spectrum. Energy and bandwidth are needed to transmit bits, and bandwidth affects capacity the most. Methods have been developed that are ten times as energy efficient at a given bandwidth consumption as simple methods. These employ signals with very complex patterns and are called "coding" solutions. The book begins with classical theory before introducing new techniques that combine older methods of error correction coding and radio transmission in order to create narrowband methods that are as efficient in both spectrum and energy as nature allows. Other topics covered include modulation techniques such as CPM, coded QAM and pulse design.

  20. OCA Code Enforcement

    Data.gov (United States)

    Montgomery County of Maryland — The Office of the County Attorney (OCA) processes Code Violation Citations issued by County agencies. The citations can be viewed by issued department, issued date...