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Sample records for ramps high burnup

  1. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  2. 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

  3. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  4. A high voltage programmable ramp generator

    Science.gov (United States)

    Upadhyay, J.; Joshi, M. J.; Deshpande, P. P.; Sharma, M. L.; Navathe, C. P.

    2008-05-01

    In this paper, a ramp generator with programmable slope is presented. It consists of a high voltage step generator, followed by integrator. The capacitor and inductor in the integrator are designed such that they can be varied by a microcontroller. This circuit generates two bipolar ramps with fastest speed <1ns and provides continuous speed variation from 6to30ns for a ramp of 500V. This is being developed as a part of automated streak camera for deflection of electron beam.

  5. Power excursion analysis for BWR`s at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  6. Effect of burn-up and high burn-up structure on spent nuclear fuel alteration

    Energy Technology Data Exchange (ETDEWEB)

    Clarens, F.; Gonzalez-Robles, E.; Gimenez, F. J.; Casas, I.; Pablo, J. de; Serrano, D.; Wegen, D.; Glatz, J. P.; Martinez-Esparza, A.

    2009-07-01

    In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    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 {sup 235}U 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 {sup 233}U, {sup 242}Pu, {sup 150}Nd, and {sup 133}Cs 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.

  8. A high burnup model developed for the DIONISIO code

    Energy Technology Data Exchange (ETDEWEB)

    Soba, A. [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, A., E-mail: denis@cnea.gov.ar [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Romero, L. [U.A. Reactores Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Villarino, E.; Sardella, F. [Departamento Ingeniería Nuclear, INVAP SE, Comandante Luis Piedra Buena 4950, 8430 San Carlos de Bariloche, Río Negro (Argentina)

    2013-02-15

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO{sub 2} fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in {sup 235}U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  9. A high burnup model developed for the DIONISIO code

    Science.gov (United States)

    Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.

    2013-02-01

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  10. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  11. Thermodynamic analysis for high burn-up fuel internal chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan). Faculty of Science and Technology

    1997-09-01

    Chemical states of fission products and actinide elements in high burn-up LWR fuel pellets have been analyzed thermodynamically using the computer program SOLGASMIX-PV. Calculations with this computer code have been performed for a complex multi-component system, which comprises 54 chemical species. The analysis shows that neither alkali nor alkaline-earth uranates are formed, but alkali and alkaline-earth molybdates exist in irradiated LWR fuel pellets in contrast with their post irradiation examinations. These molybdates tend to increase with increasing oxygen potential in the fuel under operating conditions, whereas the zirconates decrease. (author)

  12. 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).

  13. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. 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.

  16. 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.

  17. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

  18. 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.

  19. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  20. New results from the NSRR experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  1. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  2. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W. [OECD Halden Reactor Project (Norway)

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  3. 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.

  4. 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.

  5. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  6. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  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. 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.

  9. 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.

  10. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  11. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  12. High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

    Science.gov (United States)

    Lemoine, F.

    1997-09-01

    Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

  13. 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.

  14. Extension and validation of the TRANSURANUS burn-up model for helium production in high burn-up LWR fuels

    Science.gov (United States)

    Botazzoli, Pietro; Luzzi, Lelio; Brémier, Stephane; Schubert, Arndt; Van Uffelen, Paul; Walker, Clive T.; Haeck, Wim; Goll, Wolfgang

    2011-12-01

    The TRANSURANUS burn-up model (TUBRNP) calculates the local concentration of the actinides, the main fission products, and 4He as a function of the radial position across a fuel rod. In this paper, the improvements in the helium production model as well as the extensions in the simulation of 238-242Pu, 241Am, 243Am and 242-245Cm isotopes are described. Experimental data used for the extended validation include new EPMA measurements of the local concentrations of Nd and Pu and recent SIMS measurements of the radial distributions of Pu, Am and Cm isotopes, both in a 3.5% enriched commercial PWR UO 2 fuel with a burn-up of 80 and 65 MWd/kgHM, respectively. Good agreement has been found between TUBRNP and the experimental data. The analysis has been complemented by detailed neutron transport calculations (VESTA code), and also revealed the need to update the branching ratio for the 241Am(n,γ) 242mAm reaction in typical PWR conditions.

  15. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Bang, Je Geon; Kim, Dae Ho; Yang, Yong Sik; Song, Keun Woo

    2007-12-15

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced.

  16. 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.

  17. Development of the CANDU high-burnup fuel design/analysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

    1997-08-01

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs.

  18. 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.

  19. 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.

  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. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    Science.gov (United States)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  2. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  3. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  4. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

  5. 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.

  6. Oxygen potential in the rim region of high burnup UO 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  7. An empirical formulation to describe the evolution of the high burnup structure

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-15

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  8. An empirical formulation to describe the evolution of the high burnup structure

    Science.gov (United States)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  9. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  10. 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).

  11. Thermodynamic analysis for high burn-up fuel internal chemistry. 2

    Energy Technology Data Exchange (ETDEWEB)

    Fuji, Kensho; Kyoh, Bunkei [Kinki Univ., Higashi-Osaka, Osaka (Japan)

    1998-09-01

    Thermodynamic calculations with the computer program SOLGASMIX-PV have been performed for the chemical states expected in irradiated fast breeder reactor (FBR) fuels containing transuranium (TRU) elements. The analysis shows that A (alkali and alkaline-earth)-molybdates exist, but neither A-uranates nor A-zirconates are formed in FBR fuel pellets irradiated to high burn-up. And increase of oxygen potential in irradiated FBR fuel is ascribed to growing amount of rare earth, noble metal and TRU elements. (author)

  12. EBSD and TEM characterization of high burn-up mixed oxide fuel

    Science.gov (United States)

    Teague, Melissa; Gorman, Brian; Miller, Brandon; King, Jeffrey

    2014-01-01

    Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to ∼1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had ∼2.5× higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice ∼25 μm cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

  13. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Science.gov (United States)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  14. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  15. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

  16. 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).

  17. 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.

  18. 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

    ...: U.S. Department of Energy, C/O Melissa Bates, 1955 Freemont Ave., MS 1235, Idaho Falls, ID 83415..., 1955 Fremont Ave., Attn: Melissa Bates, Idaho Falls, ID, between 8 a.m. and 3:30 p.m. MT, Monday.... Melissa Bates, Contracting Officers Representative, High Burnup Dry Storage Cask Research and...

  19. High burnup effects on fuel behaviour under accident conditions: the tests CABRI REP-Na

    Science.gov (United States)

    Schmitz, Franz; Papin, Joelle

    A large, performance based, knowledge and experience in the field of nuclear fuel behaviour is available for nominal operation conditions. The database is continuously completed and precursor assembly irradiations are performed for testing of new materials and innovative designs. This procedure produces data and arguments to extend licencing limits in the permanent research for economic competitiveness. A similar effort must be devoted to the establishment of a database for fuel behaviour under off-normal and accident conditions. In particular, special attention must be given to the so-called design-basis-accident (DBA) conditions. Safety criteria are formulated for these situations and must be respected without consideration of the occurrence probability and the risk associated to the accident situation. The introduction of MOX fuel into the cores of light water reactors and the steadily increasing goal burnup of the fuel call for research work, both experimental and analytical, in the field of fuel response to DBA conditions. In 1992, a significant programme step, CABRI REP-Na, has been launched by the French Nuclear Safety and Protection Institute (IPSN) in the field of the reactivity initiated accident (RIA). After performing the nine experiments of the initial test matrix it can be concluded that important new findings have been evidenced. High burnup clad corrosion and the associated degradation of the mechanical properties of the ZIRCALOY4 clad is one of the key phenomena of the fuel behaviour under accident conditions. Equally important is the evidence that transient, dynamic fission gas effects resulting from the close to adiabatic heating introduces a new explosive loading mechanism which may lead to clad rupture under RIA conditions, especially in the case of heterogeneous MOX fuel.

  20. 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%.

  1. 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.

  2. 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.

  3. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  4. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  5. Cladding stress during extended storage of high burnup spent nuclear fuel

    Science.gov (United States)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  6. Irradiation characteristics examination technology development of irradiated nuclear material and high burn-up fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Choo, Y. S.; Oh, Y. W. [and others

    2002-12-01

    The research and development for the first year of the project are performed through specialization of researchers, information from aborad and international cooperation, securement of advanced nuclear technology, development and installation of test equipment, application of external man-power, establishment of advanced test techniques, and certified test method. 1. Absolute efficiency measurement examination technology development of gamma scanning system 2. Sample preparation technology development of SEM and EPMA for micro-structural observation and chemical composition analysis 3. Irradiated high burn-up nuclear fuel transportation and test for PWR 4. Development of hot cell examination techniques and equipment 5. Acquirement of KOLAS system. In addition to the project, the following activities are carried out as follows; - PIE of Hanaro fuel(KH99H-001) - PIE of U-Mo advanced nuclear fuel irradiated at Hanaro - PIE of Hi-MET advanced nuclear fuel irradiated at Hanaro - PIE of DUPIC project - Hot cell examination of Hanaro irradiated capsule - Leaching test of PWR fuels - Surveillance test of PWR vessels - Mechanical test of CANDU pressure tubes.

  7. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...... uses an empirical gas release model combined with a strongly burn-up dependent correction term, developed by the US Nuclear Regulatory Commission. The paper presents the experimental results and the code calculations. It is concluded that the model predictions are in reasonable agreement (within 15...

  8. Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel

    Energy Technology Data Exchange (ETDEWEB)

    Asmolov, V.; Yegorova, L. [Russian Research Centre, Moscow (Russian Federation)

    1996-03-01

    Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.

  9. Extrapolating power-ramp performance criteria for current and advanced CANDU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Chassie, G.G

    2000-06-01

    To improve the precision and accuracy of power-ramp performance criteria for high-burnup fuel, we have examined in-reactor fuel performance data as well as out-reactor test data. The data are consistent with some of the concepts used in the current formulations for defining fuel failure thresholds, such as size of power-ramp and extent of burnup. Our review indicates that there is a need to modify some other aspects of the current formulations; therefore, a modified formulation is presented in this paper. The improvements mainly concern corrodent concentration and its relationships with threshold stress for failure. The new formulation is consistent with known and expected trends such as strength of Zircaloy in corrosive environment, timing of the release of fission products to the pellet-to-sheath gap, CANLUB coating, and fuel burnup. Because of the increased precision and accuracy, the new formulation is better able to identify operational regimes that are at risk of power-ramp failures; this predictive ability provides enhanced protection to fuel against power-ramp defects. At die same time, by removing unnecessary conservatisms in other areas, the new formulation permits a greater range of defect-free operational envelope as well as larger operating margins in regions that are, in fact, not prone to power-ramp failures. (author)

  10. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    Science.gov (United States)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    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 UO2 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 UO2. 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 UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 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 UO2 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 UO2. 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.

  11. 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.

  12. 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

  13. REDUCTION IN PROBABILITY OF TRAFFIC CONGESTION ON HIGH-CLASS ROAD USING RAMP ACCESS CONTROL

    Directory of Open Access Journals (Sweden)

    R. Yu. Lagerev

    2016-01-01

    Full Text Available Мerging traffic junctions on high-class roads are considered as bottlenecks in the network and quality of their operation determines a probability for formation of traffic congestions. Investigations on congestion situations in the merging zones of ramp and freeway traffic flows have demonstrated that queuing ramp traffic flow leads to formation of so called “turbulence” effect due to re-arrangement of transport facilities and reduction in their speed on main road direction. Having high queuing traffic flow on main road the “turbulence” component can result in formation of an impact blow in the main traffic flow. It has been proved that an impact of the ramp traffic flow on congestion probability is higher in comparison with main road traffic flow. The paper makes it possible to establish that some transport facilities moving along a high-way simul taneously occupy two lanes in the merging traffic zones and they reduce capacity of the used road section. It is necessary to take into account this specific feature and it is necessary to pay attention to it in the zones of “turbulence” effect formation. The paper presents main approaches, methodology, principles and stages required for access control of high-class roads which are directed on higher quality of their operation including improvement of road traffic safety. The paper proposes a methodоlogy that allows to evaluate and optimize ramp control in the context of a transport queue length minimization at adjoining ramps for the purposes of probability reduction in transport congestion.

  14. 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.

  15. 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.

  16. Burn-Up Determination by High Resolution Gamma Spectrometry: Axial and Diametral Scanning Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R.S.; Blackadder, W.H.; Ronqvist, N.

    1967-02-15

    In the gamma spectrometric determination of burn-up the use of a single fission product as a monitor of the specimen fission rate is subject to errors caused by activity saturation or, in certain cases, fission product migration. Results are presented of experiments in which all the resolvable gamma peaks in the fission product spectrum have been used to calculate the fission rate; these results form a pattern which reflect errors in the literature values of the gamma branching ratios, fission yields etc., and also represent a series of empirical correction factors. Axial and diametral scanning experiments on a long-irradiated low-enrichment fuel element are also described and demonstrate that it is possible to differentiate between fissions in U-235 and in Pu-239 respectively by means of the ratios of the Ru-106 activity to the activities of the other fission products.

  17. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    Science.gov (United States)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  18. 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

  19. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10{sup 25} m{sup -2}, respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  20. 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.

  1. 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.

  2. 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

  3. 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

  4. High ramp rate thermogravimetric analysis of zirconium(II) hydride and titanium(II) hydride

    Energy Technology Data Exchange (ETDEWEB)

    Licavoli, Joseph J., E-mail: jjlicavo@mtu.edu; Sanders, Paul G., E-mail: sanders@mtu.edu

    2015-09-20

    Highlights: • A unique arc image device has been proposed for high ramp rate thermogravimetry. • Powder oxidation influences decomposition kinetics at temperatures below 933 K. • Particle size has a negligible effect on TiH{sub 2} decomposition behavior. • Improvements to the device are required to conduct accurate kinetic analysis. - Abstract: Zirconium and titanium hydride are utilized in liquid phase metal foam processing techniques. This application results in immediate exposure to molten metal and almost immediate decomposition at high temperatures. Most decomposition characterization techniques utilize slow heating rates and are unable to capture the decomposition behavior of hydrides under foam processing conditions. In order to address this issue a specialized high ramp rate thermogravimetric analyzer was created from a xenon arc image refiner. In addition to thermogravimetry, complimentary techniques including X-ray diffraction and scanning electron microscopy were used to characterize hydride decomposition and compare the results to literature. Hydrides were partially oxidized and separated into particles size ranges to evaluate the influence of these factors on decomposition. Oxidizing treatments were found to decrease decomposition rate only at temperatures below 933 K (660 °C) while particle size effects appeared to be negligible. Several improvements to the unique TGA apparatus presented in the current work are suggested to allow reliable kinetic modeling and analysis.

  5. Determinants of VO(2) kinetics at high power outputs during a ramp exercise protocol.

    Science.gov (United States)

    Lucía, Alejandro; Rivero, José-Luis L; Pérez, Margarita; Serrano, Antonio L; Calbet, José A L; Santalla, Alfredo; Chicharro, José L

    2002-02-01

    To determine the relationship between the additional, nonlinear increase in oxygen uptake (Delta VO(2)) that occurs at high power outputs during a ramp cycle ergometer test, on one hand; and possible explanatory mechanisms of the phenomenon, such as cardiorespiratory work, blood lactate, fitness level, or muscle fiber distribution, on the other. Ten healthy, sedentary young adults (age (mean +/- SEM), 22 +/- 1 yr) were chosen as subjects. A muscle biopsy specimen was taken from the vastus lateralis of the right leg to determine fiber type distribution by immunohistochemical identification of myosin heavy chain (MHC) isoforms. During the ramp tests (power output increases of 5 W every 15-s interval), the ventilatory threshold (VT) and lactate threshold (LT) were measured. We defined Delta VO(2) as the difference between "true" VO(2) values observed at the maximal power output (VO(2)obs) and those expected (VO(2)exp) from the previous linear VO2:power output relationship below the VT. A nonlinear increase was observed in VO2 (Delta VO(2) = 239 +/- 79 mL x min(-1), P < 0.05 for VO(2)obs vs VO(2)exp), which was significantly correlated with the percentage of type IIX fibers (r = 0.80, P < 0.05). No other correlations were found between Delta VO(2) and possible explanatory mechanisms. A greater percentage of type IIX fibers is associated with a higher excess VO(2) at high power outputs (above VT).

  6. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  7. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

  8. Raman micro-spectroscopy of UOX and MOX spent nuclear fuel characterization and oxidation resistance of the high burn-up structure

    Science.gov (United States)

    Jegou, C.; Gennisson, M.; Peuget, S.; Desgranges, L.; Guimbretière, G.; Magnin, M.; Talip, Z.; Simon, P.

    2015-03-01

    Raman micro-spectroscopy was applied to study the structure and oxidation resistance of UO2 (burnup 60 GWd/tHM) and MOX (burnup 47 GWd/tHM) irradiated fuels. The Raman technique, adapted to working under extreme conditions, enabled structural information to be obtained at the cubic micrometer scale in various zones of interest within irradiated fuel (central and zones like the Rim for UOX60, and the plutonium-enriched agglomerates for MOX47 characterized by a high burn-up structure), and the study of their oxidation resistance. As regards the structural information after irradiation, the spectra obtained make up a set of data consistent with the systematic presence of the T2g band characteristic of the fluorite structure, and of a triplet band located between 500 and 700 cm-1. The existence of this triplet can be attributed to the presence of defects originating in changes to the fuel chemistry occurring in the reactor (presence of fission products) and to the accumulation of irradiation damage. As concerns the oxidation resistance of the different zones of interest, Raman spectroscopy results confirmed the good stability of the restructured zones (plutonium-enriched agglomerates and Rim) rich in fission products compared to the non-restructured UO2 grains. A greater structural stability was noticed in the case of high plutonium content agglomerates, as this element favors the maintenance of the fluorite structure.

  9. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  10. 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)

  11. 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.

  12. Fabrications aspects of microwave devices, including ramp-type high-Tc Josephson junctions and log-periodic antenna's

    NARCIS (Netherlands)

    Terpstra, D.; Rijnders, A.J.H.M.; Roesthuis, F.J.G.; Blank, D.H.A.; Gerritsma, G.J.; Rogalla, H.

    1993-01-01

    We describe the development of high-Tc Josephson junction devices for applications at millimeter wave frequencies. These devices consist of ramp type YBCO/PBCO/YBCO Josephson junctions that are equipped with a noble metal log-periodic antenna. Growth conditions of all layers, as well as etching, cle

  13. An attempt to reproduce high burn-up structure by ion irradiation of SIMFUEL

    Energy Technology Data Exchange (ETDEWEB)

    Baranov, V.G. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation); Lunev, A.V., E-mail: AVLunev@mephi.ru [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation); Reutov, V.F. [Joint Institute for Nuclear Research (JINR), Flerov Laboratory of Nuclear Reactions (FLNR), 141980 Dubna, Moscow Region (Russian Federation); Tenishev, A.V.; Isaenkova, M.G.; Khlunov, A.V. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye Shosse 31, Moscow 115409 (Russian Federation)

    2014-09-15

    Experiments in IC-100 and U-400 cyclotrons were conducted with SIMFUEL pellets (11.47 wt.% of fission products simulators) to reproduce some aspects of the long-term irradiation conditions in epithermal reactors. Pellets were irradiated with Xe{sup 16+}, Xe{sup 24+} and He{sup +} at energies ranging from 20 keV (He{sup +}) to 320 keV (Xe{sup 16+}) and 1–90 MeV (Xe{sup 24+}). Some samples were subsequently annealed to obtain larger grain sizes and to study defects recovery. The major microstructural changes consisted in grain sub-division observed on SEM and AFM images and change in composition registered by EPMA (pellets irradiated with 1–90 MeV Xe{sup 24+} ions at fluence of 5 × 10{sup 15} cm{sup −2}). Lattice distortion and increase in dislocation density is also noted according to X-ray data. At low energies and high fluences formation of bubbles (20 keV He{sup +} at 5.5 × 10{sup 17} cm{sup −2}) was observed. Grain sub-division exhibits full coverage of the grain body and preservation of former grain boundaries. The size of sub-grains depends on local dislocation density and changes from 200 nm to 400 nm along the irradiated surface. Beneath it the size ranges from 150 to 600 nm. Sub-grains are not observed in samples irradiated by low-energy ions even at high dislocation densities.

  14. 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.

  15. Analysis of stationary fuel cell dynamic ramping capabilities and ultra capacitor energy storage using high resolution demand data

    Science.gov (United States)

    Meacham, James R.; Jabbari, Faryar; Brouwer, Jacob; Mauzey, Josh L.; Samuelsen, G. Scott

    Current high temperature fuel cell (HTFC) systems used for stationary power applications (in the 200-300 kW size range) have very limited dynamic load following capability or are simply base load devices. Considering the economics of existing electric utility rate structures, there is little incentive to increase HTFC ramping capability beyond 1 kWs -1 (0.4% s -1). However, in order to ease concerns about grid instabilities from utility companies and increase market adoption, HTFC systems will have to increase their ramping abilities, and will likely have to incorporate electrical energy storage (EES). Because batteries have low power densities and limited lifetimes in highly cyclic applications, ultra capacitors may be the EES medium of choice. The current analyses show that, because ultra capacitors have a very low energy storage density, their integration with HTFC systems may not be feasible unless the fuel cell has a ramp rate approaching 10 kWs -1 (4% s -1) when using a worst-case design analysis. This requirement for fast dynamic load response characteristics can be reduced to 1 kWs -1 by utilizing high resolution demand data to properly size ultra capacitor systems and through demand management techniques that reduce load volatility.

  16. High Voltage Ramp Generator for Electro-Optically Tunable Filter for the MSE-CIF Diagnostics on NSTX.

    Science.gov (United States)

    Wu, Ying; Levinton, Fred

    2004-11-01

    The motional Stark effect (MSE) diagnostic is routinely used to determine the q-profile in large fusion devices. To apply the MSE diagnostic to experiments with low magnetic fields such as NSTX (<1 T), a tunable birefringent Lyot filter is used with high throughput and high resolution which allows for a good signal-to-noise ratio. The birefringent filter is made from lithium-niobate crystals, which are coated with a layer of indium tin-oxide (ITO). The ITO layer is a transparent conductive coating. By applying an electric field across the crystal the index of refraction is varied. This allows tunability of the filter. Putting multiple crystals together and tuning them individually it is possible to pass certain wavelengths of light and reject others. A high voltage ramp generator circuit is under development to ramp a 5 kV signal using a simple design involving MOSFET ladders. The goal is to design the circuit so that it can ramp ±5000 volts at a frequency of around 1 kHz. This would allow the filter to sweep over a range of ˜ 1nm.

  17. Lattice parameter changes associated with the rim-structure formation in high burn-up UO 2 fuels by micro X-ray diffraction

    Science.gov (United States)

    Spino, J.; Papaioannou, D.

    2000-10-01

    Radial variations of the lattice parameter and peak width of two high burn-up UO 2-fuels (67 and 80 GWd/tM) were measured by a specially developed micro-X-ray diffraction technique, allowing spectra acquisition with 30 μm spatial resolution. The results showed a significant but constant peak broadening, and a lattice parameter that increased towards the pellet edge and decreased again within the rim-zone. This lattice contraction coincided with other property changes in the rim region, i.e., porosity increase, hardness decrease and Xe depletion. In terms of local burn-ups, the lattice contraction followed the rate of the matrix Xe depletion measured by EMPA, exceeding greatly the contraction rate due to dissolved fission products. The observed behaviour can be equally explained by a saturation of single interstitials with subsequent recombination with excess vacancies, as by the saturation and enlargement of dislocation loops. The concentration and sizes of defects involved and their possible relation to the rim structure formation are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. 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.

  20. A PWR Thorium Pin Cell Burnup Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.

    2000-05-01

    As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.

  1. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  2. Ramping turn-to-turn loss and magnetization loss of a No-Insulation (RE)Ba2Cu3Ox high temperature superconductor pancake coil

    Science.gov (United States)

    Wang, Y.; Song, H.; Yuan, W.; Jin, Z.; Hong, Z.

    2017-03-01

    This paper is to study ramping turn-to-turn loss and magnetization loss of a no-insulation (NI) high temperature superconductor (HTS) pancake coil wound with (RE)Ba2Cu3Ox (REBCO) conductors. For insulated (INS) HTS coils, a magnetization loss occurs on superconducting layers during a ramping operation. For the NI HTS coil, additional loss is generated by the "bypassing" current on the turn-to-turn metallic contacts, which is called "turn-to-turn loss" in this study. Therefore, the NI coil's ramping loss is much different from that of the INS coil, but few studies have been reported on this aspect. To analyze the ramping losses of NI coils, a numerical method is developed by coupling an equivalent circuit network model and a H-formulation finite element method model. The former model is to calculate NI coil's current distribution and turn-to-turn loss, and the latter model is to calculate the magnetization loss. A test NI pancake coil is wound with REBCO tapes and the reliability of this model is validated by experiments. Then the characteristics of the NI coil's ramping losses are studied using this coupling model. Results show that the turn-to-turn loss is much higher than the magnetization loss. The NI coil's total ramping loss is much higher than that of its insulated counterpart, which has to be considered carefully in the design and operation of NI applications. This paper also discusses the possibility to reduce NI coil's ramping loss by decreasing the ramping rate of power supply or increasing the coil's turn-to-turn resistivity.

  3. The effect of dissolved hydrogen on the dissolution of {sup 233}U doped UO{sub 2}(s) high burn-up spent fuel and MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Carbol, P. [Inst. for Transuranium Elements, Karlsruhe (Germany); Spahiu, K. (ed.) [and others

    2005-03-01

    In this report the results of the experimental work carried out in a large EU-research project (SFS, 2001-2004) on spent fuel stability in the presence of various amounts of near field hydrogen are presented. Studies of the dissolution of {sup 233}U doped UO{sub 2}(s) simulating 'old' spent fuel were carried out as static leaching tests, autoclave tests with various hydrogen concentrations and electrochemical tests. The results of the leaching behaviour of a high burn-up spent fuel pellet in 5 M NaCl solutions in the presence of 3.2 bar H{sub 2} pressure and of MOX fuel in dilute synthetic groundwater under 53 bar H{sub 2} pressure are also presented. In all the experimental studies carried out in this project, a considerable effect of hydrogen in the dissolution rates of radioactive materials was observed. The experimental results obtained in this project with a-doped UO{sub 2}, high burn-up spent fuel and MOX fuel together with literature data give a reliable background to use fractional alteration/dissolution rates for spent fuel of the order of 10{sup -6}/yr - 10{sup -8}/yr with a recommended value of 4x10{sup -7}/yr for dissolved hydrogen concentrations above 10{sup -3} M and Fe(II) concentrations typical for European repository concepts. Finally, based on a review of the experimental data and available literature data, potential mechanisms of the hydrogen effect are also discussed. The work reported in this document was performed as part of the Project SFS of the European Commission 5th Framework Programme under contract no FIKW-CT-2001-20192 SFS. It represents the deliverable D10 of the experimental work package 'Key experiments using a-doped UO{sub 2} and real spent fuel', coordinated by SKB with the participation of ITU, FZK-INE, ENRESA, CIEMAT, ARMINES-SUBATECH and SKB.

  4. Laser-Driven Ramp Compression to Investigate and Model Dynamic Response of Iron at High Strain Rates

    Directory of Open Access Journals (Sweden)

    Nourou Amadou

    2016-12-01

    Full Text Available Efficient laser shock processing of materials requires a good characterization of their dynamic response to pulsed compression, and predictive numerical models to simulate the thermomechanical processes governing this response. Due to the extremely high strain rates involved, the kinetics of these processes should be accounted for. In this paper, we present an experimental investigation of the dynamic behavior of iron under laser driven ramp loading, then we compare the results to the predictions of a constitutive model including viscoplasticity and a thermodynamically consistent description of the bcc to hcp phase transformation expected near 13 GPa. Both processes are shown to affect wave propagation and pressure decay, and the influence of the kinetics of the phase transformation on the velocity records is discussed in details.

  5. Experiment on the improvement of sinterability for dry recycling nuclear fuel pellets by using simulated spent PWR fuel of high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, G. I.; Lee, Jae W.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Lee, J. W.; Yang, M. S.; Shin, W. C

    2004-09-01

    To study the fabrication characteristics of dry recycling nuclear fuel using spent PWR fuel with high burnup of 60,000 MWd/tU, the fission products of spent PWR fuel was analyzed by ORIGEN-2 code. Simulated spent PWR fuel pellets were fabricated by using UO{sub 2} powder added by the simulated fission products. The simulated dry-recycling-fuel pellets were fabricated by dry recycling fuel fabrication flow including 3 cycle treated OREOX(Oxidation and REduction of OXide fuel) process. A small amount of dopant such as TiO{sub 2}, Nb{sub 2}O{sub 5}, Li{sub 2}O are added to increase sinterability of the OREOX treated powder. As the results of experiments, the densities of sintered pellets without dopant ranged from 10.04 to 10.34 g/cm{sup 3}(94.3 to 97.1% of T.D.), the grain size of the pellets ranged from 3 to 4 {mu}m. The sintered density of the pellets with TiO{sub 2} or Nb{sub 2}O{sub 5} ranged from 10.46 to 10.32 g/cm{sup 3}(98.2 to 96.9 % of T.D.) The grain size of the pellets with TiO{sub 2}, Nb{sub 2}O{sub 5} or Li{sub 2}O ranged from 7.3 to 12.2 {mu}m.

  6. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    Science.gov (United States)

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  7. Calibration of burnup monitor in the Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oheda, K.; Naito, H.; Hirota, M. [Japan Nuclear Fuel Ltd., Aomori (Japan); Natsume, K. [Toshiba Corp., Yokohama, Kawasaki, Kanagawa (Japan); Kumanomido, H. [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    1998-07-01

    The Rokkasho Reprocessing Plant has adopted a credit for burnup in criticality control in the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. The burnup monitor system, prepared for BWR and PWR type fuel assemblies, nondestructively measures the burnup value and determines the residual U-235 enrichment in a spent fuel assembly, and criticality is controlled by the value of residual U-235 enrichment in SFSF and by the value of top 50 cm average burnup in the Dissolution Facility. The burnup monitor consists of three measurement systems; a Boss gamma-ray profile measurement system, a high resolution gamma-ray spectrometry system, and a passive neutron measurement system. The monitor sensitivity is calibrated against operator-declared burnup values through repetitive measurements of 100 spent fuel assemblies: BWR 8 X 8, PWR 14 X 14. and 17 X 17. The outline of the measurement methods, objectives of the calibration, actual calibration method, and an example of calibration performed in a demonstration experiment are presented. (author)

  8. Heterogeneous UO2 fuel irradiated up to a high burn-up: Investigation of the HBS and of fission product releases

    Science.gov (United States)

    Noirot, J.; Lamontagne, J.; Nakae, N.; Kitagawa, T.; Kosaka, Y.; Tverberg, T.

    2013-11-01

    A UO2 fuel with a heterogeneous distribution of 235U was irradiated up to a high burn-up in the Halden Boiling Water Reactor (HBWR). The last 100 days of irradiation were performed with an increased level of linear power. The effect of the heterogeneous fissile isotope distribution on the formation of the HBS was studied free of the possible influence of Pu which exists in heterogeneous MOX fuels. The HBS formed in 235U-rich agglomerates and its main characteristics were very similar to those of the HBS formed in Pu-rich agglomerates of heterogeneous MOX fuels. The maximum local contents of Nd and Xe before HBS formation were studied in this fuel. In addition to a Pu effect that promotes the HBS phenomenon, comparison with previous results for heterogeneous MOX fuels showed that the local fission product concentration was not the only parameter that has to be taken into consideration. It appears that the local actinide depletion by fission and/or the energy locally deposited through electronic interactions in the fission fragment recoils also have an effect on the HBS formation threshold. Moreover, a major release of fission gases from the peripheral 235U-rich agglomerates of HBS bubbles and a Cs radial movement are also evidenced in this heterogeneous UO2. Cs deposits on the peripheral grain boundaries, including the HBS grain boundaries, are considered to reveal the release paths. SUP>235U-rich agglomerates, SUP>235U-poor areas, an intermediate phase with intermediate 235U concentrations. Short fuel rods were fabricated with these pellets. The main characteristics of these fuel rods are shown in Table 1.These rods were irradiated to high burn-ups in the IFA-609/626 of the HBWR and then one was irradiated in the IFA-702 for 100 days. Fig. 2 shows the irradiation history of this fuel. The final average burn-up of the rod was 69 GWd/tU. Due to the flux differences along the rod, however, the average burn-up of the cross section examined was 63 GWd/tU. This fuel

  9. Development of Burnup Calculation Code for Pebble-bed High Temperature Reactor at Equilibrium State%球床高温堆平衡态燃耗计算程序的开发

    Institute of Scientific and Technical Information of China (English)

    朱贵凤; 邹杨; 李明海; 严睿; 彭红花; 徐洪杰

    2015-01-01

    The burnup calculation code PBRE coupling MCNP5 and ORIGEN2 was developed for pebble‐bed high temperature reactor at equilibrium state ,and it can be used to analyze the neutronic performance of equilibrium core .The iteration method was optimized in order to save Monte Carlo calculation time ,and the convergence can be reached in 10 iterative steps .The average discharged burnup for HTR‐10 is consistent with literature ,and it indicates that the PBRE is suitable to analyze the burnup for pebble‐bed reactor at equilibrium state .%基于MCNP5和ORIGEN2耦合方法,开发了平衡态下球床高温堆的燃耗计算程序PBRE ,用于堆的性能价值分析。为节省蒙特卡罗计算时间,对迭代收敛的方法进行优化,使之可在10个迭代步内收敛。使用PBRE对清华大学H T R‐10进行建模计算,得到的平均卸料燃耗深度与文献报道值一致,表明PBRE程序适用于球床堆平衡态的燃耗分析。

  10. Chemical states of fission products in irradiated (U{sub 0.3}Pu{sub 0.7})C{sub 1+x} fuel at high burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Renu [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: arenu@barc.gov.in; Venugopal, V. [Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel (U{sub 0.3}Pu{sub 0.7})C{sub 1+x}, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  11. A self-focusing, high transformer ratio, collinear plasma dielectric wakefield accelerator driven by a ramped bunch train

    Science.gov (United States)

    Sotnikov, Gennadij V.; Marshall, Thomas C.; Shchelkunov, Sergey V.; Hirshfield, Jay L.

    2017-03-01

    New results of studies of wakefield excitation by a ramped bunch train in a collinear, single-channel dielectriclined THz-wakefield accelerator structure that is filled with a low-temperature plasma are presented. A novel ramped train of drive bunches, together with plasma filling part of the transport channel, makes possible substantial improvement of the transformer ratio of the multimode collinear device to 6:1 while the plasma could stabilize the transverse motion of the drive and witness bunches.

  12. RELATION BETWEEN PORE MODEL AND CENTER-LINE TEMPERATURE IN HIGH BURN-UP UO2 PELLET

    Directory of Open Access Journals (Sweden)

    Suwardi Suwardi

    2010-06-01

    Full Text Available Relation between pore model and center-line temperature of high burn up UO2 Pellet. Temperature distribution has been evaluated by using different model of pore distribution. Typical data of power distribution and coolant data have been chosen in this study. Different core model and core distribution model have been studied for related temperature, in correlation with high burn up thermal properties. Finite element combined finite different adapted from Saturn-1 has been used for calculating the temperature distribution. The center-line temperature for different pore model and related discussion is presented.   Keywords: pore model, high burn up, UO2 pellet, centerline temperature.

  13. High-Pressure Liquid Chromatography of Irradiated Nuclear Fue - Separation of Neodymium for Burn-up Determination

    DEFF Research Database (Denmark)

    Larsen, N. R.

    1979-01-01

    Neodymium is separated from solutions of spent nuclear fuel by high-pressure liquid chromatography in methanol-nitric acid-water media using an anion-exchange column. Chromatograms obtained by monitoring at 280 nm, illustrate the difficulties especially with the fission product ruthenium in nuclear...

  14. Calibration of burnup monitor installed in Rokkasho Reprocessing Plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Naito, Hirofumi; Hirota, Masanari [Japan Nuclear Fuel Co. Ltd., Rokkasho, Aomori (Japan); Natsume, Koichiro [Isogo Engineering Center, Toshiba Corporation, Yokohama, Kanagawa (Japan); Kumanomido, Hironori [Nuclear Engineering Laboratory, Toshiba Corporation, Kawasaki, Kanagawa (Japan)

    2000-06-01

    Rokkasho Reprocessing Plant uses burnup credit for criticality control at the Spent Fuel Storage Facility (SFSF) and the Dissolution Facility. A burnup monitor measures nondestructively burnup value of a spent fuel assembly and guarantees the credit for burnup. For practical reasons, a standard radiation source is not used in calibration of the burnup monitor, but the burnup values of many spent fuel assemblies are measured based on operator-declared burnup values. This paper describes the concept of burnup credit, the burnup monitor, and the calibration method. It is concluded, from the results of calibration tests, that the calibration method is valid. (author)

  15. Fabrication characteristics of dry process fuel with a variation of fuel burn-ups

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Kim, W. K.; Lee, J. W. [and others

    2004-11-01

    Fabrication characteristics of the dry processed fuel with a variation of fuel burn-ups in a range of 27,300 to 65,000 MWD/tU were experimentally evaluated. Density comparison of powders which were fabricated from oxidation, OREOX and milling processes at same process conditions was performed with a function of fuel burn-ups respectively. The influence of fuel burn-ups on sintering characteristics of dry processed fuel was studied by comparing the density change of sintered pellet as well as green pellet. Weight gain by fuel oxidation to U{sub 3}O{sub 8} showed semi-linear dependence with increasing fuel burn-ups. OREOX powder density increased up to 3.7 g/cm{sup 3} at high burn-up fuel, and the density of milled powder with fuel burn-ups represented almost similar value of 3.2{+-}0.2 g/cm{sup 3}. Also, the green pellet density compacted by 120 MPa decreased smoothly with increasing fuel burn-ups, and the density change of sintered pellet showed the similar trend as green pellet. The sintered density of pellet in a range of 27,000 to 40,000 MWD/tU was found to be more 95% of Theoretical Density(T.D.), but the sintered pellet density fabricated from high burn-up fuel showed a range of 92 % to 93% of T.D.

  16. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  17. Strategies for Application of Isotopic Uncertainties in Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2002-12-23

    Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103

  18. RAPID program to predict radial power and burnup distribution of UO{sub 2} fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Song, Jae Sung; Bang, Je Gun; Kim, Dae Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    Due to the radial variation of the neutron flux and its energy spectrum inside UO{sub 2} fuel, the fission density and fissile isotope production rates are varied radially in the pellet, and it becomes necessary to know the accurate radial power and burnup variation to predict the high burnup fuel behavior such as rim effects. Therefore, to predict the radial distribution of power, burnup and fissionable nuclide densities in the pellet with the burnup and U-235 enrichment, RAPID(RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet) program was developed. It considers the specific radial variation of the neutron reaction of the nuclides while the constant radial variation of neutron reaction except neutron absorption of U-238 regardless of the nuclides, the burnup and U-235 enrichment is assumed in TUBRNP model which is recognized as the one of the most reliable models. Therefore, it is expected that RAPID may be more accurate than TUBRNP, specially at high burnup region. RAPID is based upon and validated by the detailed reactor physics code, HELIOS which is one of few codes that can calculates the radial variations of the nuclides inside the pellet. Comparison of RAPID prediction with the measured data of the irradiated fuels showed very good agreement. RAPID can be used to calculate the local variations of the fissionable nuclide concentrations as well as the local power and burnup inside that pellet as a function of the burnup up to 10 w/o U-235 enrichment and 150 MWD/kgU burnup under the LWR environment. (author). 8 refs., 50 figs., 1 tab.

  19. Wind Plant Ramping Behavior

    Energy Technology Data Exchange (ETDEWEB)

    Ela, E.; Kemper, J.

    2009-12-01

    With the increasing wind penetrations, utilities and operators (ISOs) are quickly trying to understand the impacts on system operations and planning. This report focuses on ramping imapcts within the Xcel service region.

  20. Meniscal Ramp Lesions

    Science.gov (United States)

    Chahla, Jorge; Dean, Chase S.; Moatshe, Gilbert; Mitchell, Justin J.; Cram, Tyler R.; Yacuzzi, Carlos; LaPrade, Robert F.

    2016-01-01

    Meniscal ramp lesions are more frequently associated with anterior cruciate ligament (ACL) injuries than previously recognized. Some authors suggest that this entity results from disruption of the meniscotibial ligaments of the posterior horn of the medial meniscus, whereas others support the idea that it is created by a tear of the peripheral attachment of the posterior horn of the medial meniscus. Magnetic resonance imaging (MRI) scans have been reported to have a low sensitivity, and consequently, ramp lesions often go undiagnosed. Therefore, to rule out a ramp lesion, an arthroscopic evaluation with probing of the posterior horn of the medial meniscus should be performed. Several treatment options have been reported, including nonsurgical management, inside-out meniscal repair, or all-inside meniscal repair. In cases of isolated ramp lesions, a standard meniscal repair rehabilitation protocol should be followed. However, when a concomitant ACL reconstruction (ACLR) is performed, the rehabilitation should follow the designated ACLR postoperative protocol. The purpose of this article was to review the current literature regarding meniscal ramp lesions and summarize the pertinent anatomy, biomechanics, diagnostic strategies, recommended treatment options, and postoperative protocol. PMID:27504467

  1. Forward ramp in 3D

    Science.gov (United States)

    1997-01-01

    Mars Pathfinder's forward rover ramp can be seen successfully unfurled in this image, taken in stereo by the Imager for Mars Pathfinder (IMP) on Sol 3. 3D glasses are necessary to identify surface detail. This ramp was not used for the deployment of the microrover Sojourner, which occurred at the end of Sol 2. When this image was taken, Sojourner was still latched to one of the lander's petals, waiting for the command sequence that would execute its descent off of the lander's petal.The image helped Pathfinder scientists determine whether to deploy the rover using the forward or backward ramps and the nature of the first rover traverse. The metallic object at the lower left of the image is the lander's low-gain antenna. The square at the end of the ramp is one of the spacecraft's magnetic targets. Dust that accumulates on the magnetic targets will later be examined by Sojourner's Alpha Proton X-Ray Spectrometer instrument for chemical analysis. At right, a lander petal is visible.The IMP is a stereo imaging system with color capability provided by 24 selectable filters -- twelve filters per 'eye.' It stands 1.8 meters above the Martian surface, and has a resolution of two millimeters at a range of two meters.Mars Pathfinder is the second in NASA's Discovery program of low-cost spacecraft with highly focused science goals. The Jet Propulsion Laboratory, Pasadena, CA, developed and manages the Mars Pathfinder mission for NASA's Office of Space Science, Washington, D.C. JPL is an operating division of the California Institute of Technology (Caltech). The Imager for Mars Pathfinder (IMP) was developed by the University of Arizona Lunar and Planetary Laboratory under contract to JPL. Peter Smith is the Principal Investigator.Click below to see the left and right views individually. [figure removed for brevity, see original site] Left [figure removed for brevity, see original site] Right

  2. Fission-gas release at extended burnups: effect of two-dimensional heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Yu, S.D. [Ryerson Polytechnic Univ., Toronto, Ontario (Canada); Lau, J.H.K

    2000-09-01

    To better simulate the performance of high-burnup CANDU fuel, a two-dimensional model for heat transfer between the pellet and the sheath has been added to the computer code ELESTRES. The model covers four relative orientations of the pellet and the sheath and their impacts on heat transfer and fission-gas release. The predictions of the code were compared to a database of 27 experimental irradiations involving extended burnups and normal burnups. The calculated values of fission gas release matched the measurements to an average of 94%. Thus, the two-dimensional heat transfer model increases the versatility of the ELESTRES code to better simulate fuels at normal as well as at extended burnups. (author)

  3. Investigation on using neutron counting techniques for online burnup monitoring of pebble bed reactor fuels

    Science.gov (United States)

    Zhao, Zhongxiang

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 104 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 1013 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble's burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ˜7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3

  4. Meniscal Ramp Lesions

    OpenAIRE

    2016-01-01

    Meniscal ramp lesions are more frequently associated with anterior cruciate ligament (ACL) injuries than previously recognized. Some authors suggest that this entity results from disruption of the meniscotibial ligaments of the posterior horn of the medial meniscus, whereas others support the idea that it is created by a tear of the peripheral attachment of the posterior horn of the medial meniscus. Magnetic resonance imaging (MRI) scans have been reported to have a low sensitivity, ...

  5. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  6. Integrated burnup calculation code system SWAT

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Hirakawa, Naohiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwasaki, Tomohiko

    1997-11-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user`s manual of SWAT. (author)

  7. Micro-Ramps for Hypersonic Flow Control

    Directory of Open Access Journals (Sweden)

    Konstantinos Kontis

    2012-04-01

    Full Text Available Shock/boundary layer interaction (SBLI is an undesirable phenomenon, occurring in high-speed propulsion systems. The conventional method to manipulate and control SBLI is using a bleed system that involves the removal of a certain amount of mass of the inlet flow to control boundary layer separation. However, the system requires a larger nacelle to compensate the mass loss, larger nacelles contribute to additional weight and drag and reduce the overall performance. This study investigates a novel type of flow control device called micro-ramps, a part of the micro vortex generators (VGs family that intends to replace the bleed technique. Micro-ramps produce pairs of counter-rotating streamwise vortices, which help to suppress SBLI and reduce the chances of flow separation. Experiments were done at Mach 5 with two micro-ramp models of different sizes. Schlieren photography, surface flow visualization and infrared thermography were used in this investigation. The results revealed the detailed flow characteristics of the micro-ramp, such as the primary and secondary vortices. This helps us to understand the overall flow physics of micro-ramps in hypersonic flow and their application for SBLI control.

  8. Post-irradiation examinations and high-temperature tests on undoped large-grain UO2 discs

    Science.gov (United States)

    Noirot, J.; Pontillon, Y.; Yagnik, S.; Turnbull, J. A.

    2015-07-01

    Within the Nuclear Fuel Industry Research (NFIR) programme, several fuel variants -in the form of thin circular discs - were irradiated in the Halden Boiling Water Reactor (HBWR) at burn-ups up to ∼100 GWd/tHM. The design of the fuel assembly was similar to that used in other HBWR programmes: the assembly contained several rods with fuel discs sandwiched between Mo discs, which limited temperature differences within each fuel disc. One such variant was made of large-grain UO2 discs (3D grain size = ∼45 μm) which were subjected to three burn-ups: 42, 72 and 96 GWd/tHM. Detailed characterizations of some of these irradiated large-grain UO2 discs were performed in the CEA Cadarache LECA-STAR hot laboratory. The techniques used included electron probe microanalysis (EPMA), scanning electron microscopy (SEM) and secondary ion mass spectrometry (SIMS). Comparisons were then carried out with more standard grain size UO2 discs irradiated under the same conditions. Examination of the high burn-up large-grain UO2 discs revealed the limited formation of a high burn-up structure (HBS) when compared with the standard-grain UO2 discs at similar burn-up. High burn-up discs were submitted to temperature transients up to 1200 °C in the heating test device called Merarg at a relatively low temperature ramp rate (0.2 °C/s). In addition to the total gas release during these tests, the release peaks throughout the temperature ramp were monitored. Tests at 1600 °C were also conducted on the 42 GWd/tHM discs. The fuels were then characterized with the same microanalysis techniques as those used before the tests, to investigate the effects of these tests on the fuel's microstructure and on the fission gas behaviour. This paper outlines the high resistance of this fuel to gas precipitation at high temperature and to HBS formation at high burn-up. It also shows the similarity of the positions, within the grains, where HBS forms at high burn-up and where bubbles appear during the low

  9. 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.)

  10. ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

    Energy Technology Data Exchange (ETDEWEB)

    A.H. Wells

    2004-11-17

    The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

  11. 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

  12. Determination of IRT-2M fuel burnup by gamma spectrometry.

    Science.gov (United States)

    Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav

    2016-01-01

    A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.

  13. MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, T.; Sternat, M.; Charlton, W.

    2011-05-08

    MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

  14. 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.

  15. Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

    Science.gov (United States)

    Fensin, Michael Lorne

    Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established

  16. A survey on wind power ramp forecasting.

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, C.; Gama, J.; Matias, L.; Botterud, A.; Wang, J. (Decision and Information Sciences); (INESC Porto)

    2011-02-23

    The increasing use of wind power as a source of electricity poses new challenges with regard to both power production and load balance in the electricity grid. This new source of energy is volatile and highly variable. The only way to integrate such power into the grid is to develop reliable and accurate wind power forecasting systems. Electricity generated from wind power can be highly variable at several different timescales: sub-hourly, hourly, daily, and seasonally. Wind energy, like other electricity sources, must be scheduled. Although wind power forecasting methods are used, the ability to predict wind plant output remains relatively low for short-term operation. Because instantaneous electrical generation and consumption must remain in balance to maintain grid stability, wind power's variability can present substantial challenges when large amounts of wind power are incorporated into a grid system. A critical issue is ramp events, which are sudden and large changes (increases or decreases) in wind power. This report presents an overview of current ramp definitions and state-of-the-art approaches in ramp event forecasting.

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

    Science.gov (United States)

    Makmal, T.; Aviv, O.; Gilad, E.

    2016-10-01

    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.

  18. Revised SWAT. The integrated burnup calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)

    2000-07-01

    SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)

  19. 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.

  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. 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.

  2. 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.

  3. A Simple Global View of Fuel Burnup

    Science.gov (United States)

    Sekimoto, Hiroshi

    2017-01-01

    Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

  4. Global Decoupling on the RHIC Ramp

    CERN Document Server

    Luo, Yun; Della Penna, Al; Fischer, Wolfram; Laster, Jonathan S; Marusic, Al; Pilat, Fulvia Caterina; Roser, Thomas; Trbojevic, Dejan

    2005-01-01

    The global betatron decoupling on the ramp is an important issue for the operation of the Relativistic Heavy Ion Collider (RHIC). In the polarized proton run, the betatron tunes are required to keep almost constant on the ramp to avoid spin resonance line crossing and the beam polarization loss. Some possible correction schemes on the ramp, like three-ramp correction, the coupling amplitude modulation and the coupling phase modulaxtion, have been found. The principles of these schemes are shortly reviewed and compared. Operational results of their applications on the RHIC ramps are given.

  5. Traffic Congestion Mechanism in Two Ramp Systems

    Institute of Scientific and Technical Information of China (English)

    ZHAO Xiao-Mei; SONG Gui-Cui; SONG Yu-Kun

    2011-01-01

    The aim of this paper is to study traffic properties in an on/off-ramp system with a bus stop close to the on/off ramp.The location of the bus stop in the on/off-ramp (thereafter downstream or upstream case) is discussed.The simulation results show that in the two ramp systems, the reasons for traffic congestions are different.In the on-ramp system, buses and cars coming from on-ramp interweave each other, while in the off-ramp system, buses interweave with cars exiting to off-ramp.Thus, in the on-ramp (off-ramp) system, the upstream (downstream) bus stop is helpful to reduce the interweaving situation.Moreover, the negative effect will disappear when the distance between the bus stop and the on/off-ramp is more than 20 cells (i.e.150 m).These qualitative findings may provide some suggestions on traffic management and optimization.

  6. An investigation into the origen of the interference generated during the measurement of the reactivity in a high burn-up reactor core%高燃耗堆芯反应性测量的干扰源研究

    Institute of Scientific and Technical Information of China (English)

    陈雄月; 吕大军; 裘希春; 韩承慈; 夏应军; 邓朝平; 张仲元

    2012-01-01

    回顾了1980年9月实验前,在反应堆噪声分析领域的技术发展概况.展示了在实验动力堆燃耗末、卸料前,用双探测器互相关频谱分析法(CCFS)测得的一组数据;和经过离线去本底拟合计算后,获得的动力学参数测量结果:αc=(144.57±2.09)s-1.介绍了数据获取过程中出现的异常情况;离线处理的方法;本底谱选定;拟合计算程序;计算结果和结论.还简要介绍了干扰源的来源及其强度计算概况.数据处理结果证明:在长期燃耗后的堆芯上应用噪声分析法,除了要克服大γ场的干扰外,还要严格消除本底中子场产生的不相关噪声干扰.%After a general review for the technical development before 1980's in the area of nuclear reactor noise analysis,a reactor dynamic parameter,ac = (144. 57 + 2. 09)s~1 , obtained through off-line background processing, is shown. The processed data is measured through double- detector cross correlation frequency spectral analysis (CCFS) for the experimental nuclear power reactor at the burn-up end in sept. 1980. This paper also presents the abnormal situations for data acquisition, the off-line data processing method,the background spectra selection for data processing and the program for the least-squares fit calculation. Here also explains how neutron background is generated and how its strength is calculated. This verifies the fact that after a long-term burn-up run, large y field must be suppressed and also more attention must be paid to the uncorrelated neutron noise from the fuel burn-up.

  7. Ramping up PCI production

    Energy Technology Data Exchange (ETDEWEB)

    Bergsma, D.; Kieftenbeld, B.; Bol, L. [Corus Strip Products IJmuiden (Netherlands). Blast Furnace Dept.

    2008-07-01

    This paper described the improvements at the Corus Strip Products IJmuiden coal preparation plant and injection facilities that have resulted in ultra high productivities and very low coke rates of the blast furnaces. Pulverized coal injection (PCI) began at the steelmaking plant 25 years ago. Since then, an increased coal preparation output has been realized each year. The capacity of the mills and the injection systems were increased from about 70 tons/h to over 200 tons/h, thereby allowing a steady decrease in coke rate per ton of hot metal with a considerable increase in hot metal production from the blast furnaces. Many modifications were in hardware, but much of the increase in productivity was due to increasing the aimed pulverized coal (PC) particle size, selecting different coals and improving process control in both milling and injection facilities. Depending on the coal variety, the specific PCI-rates could reach over 250 kg/thm. The excellent performance was made possible by improving key factors such as burden quality, burdening, cast house performance and installation reliability. The modifications also solved constraints in working pressure; uptake pulverized coal moisture content; inlet and outlet air temperature; and injection rate stability. Coal selection was based on handlability, chemical composition and economics. 1 tab., 9 figs.

  8. Triton burnup measurements in KSTAR using a neutron activation system

    Science.gov (United States)

    Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.

    2016-11-01

    Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

  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. Assesment of advanced step models for steady state Monte Carlo burnup calculations in application to prismatic HTGR

    Directory of Open Access Journals (Sweden)

    Kępisty Grzegorz

    2015-09-01

    Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.

  11. LHC Report: Intensity ramp-up

    CERN Multimedia

    Mike Lamont for the LHC Team

    2012-01-01

    The first stable beams at 4 TeV were declared on Thursday, 5 April with 3 bunches per beam. This marked the start of the intensity ramp-up, which aims to get back up to 1380 bunches per beam as quickly as is safely possible.   The next couple of days saw fills with 47, 84 and 264 bunches per beam and on Sunday, 8 April the move was made to 624 bunches. With the squeeze to 60 cm in place, 624 bunches with reasonably high bunch intensities of around 1.3 to 1.4 x1011 protons per bunch have already yielded respectable peak luminosities of up to 2.5 x1033 cm-2s-1. Following a lot of hard work during the Christmas technical stop, machine availability is very good at the moment. The ramp-up in the number of bunches is accompanied by a series of checks aimed to make sure the machine protection systems and operational procedures are in a good enough shape to safely deal with the beam intensity. 624 bunches at 4 TeV already represents an energy over 50 MJ and serious damage potential. The next few days sh...

  12. 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.

  13. 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)

    Navarro, J.; Aryaeinejad, R.; Nigg, D.W. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    The goal of this work was to perform a feasibility study and establish measurement techniques to determine the burnup of the Advanced Test Reactor (ATR) fuels at the Idaho National Laboratory (INL). Three different detectors of high purity germanium (HPGe), lanthanum bromide (LaBr{sub 3}), and high pressure xenon (HPXe) in two detection system configurations of below and above the water pool were used in this study. The last two detectors were used for the first time in fuel burnup measurements. The results showed that a better quality spectra can be achieved with the above the water pool configuration. Both short and long cooling time fuels were investigated in order to determine which measurement technique, absolute or fission product ratio, is better suited in each scenario and also to establish what type of detector should be used in each case for the best burnup measurement. The burnup and cooling time calibrations were established using experimental absolute activities or isotopic ratios and ORIGEN burnup calculations. A method was developed to do burnup and cooling time calibrations using fission isotopes activities without the need to know the exact geometry. (authors)

  14. Quantification of tsunami-induced flows on a Mediterranean carbonate ramp reveals catastrophic evolution

    NARCIS (Netherlands)

    Slootman, A.; Cartigny, M.J.B.; Moscariello, A.; Chiaradia, M.; de Boer, P.L.

    2016-01-01

    Cool-water carbonates are the dominant limestones in the Mediterranean Basin since the Early Pliocene. Their deposition typically resulted in ramp morphologies due to high rates of resedimentation. Several such fossil carbonate ramps are characterised by a bimodal facies stacking pattern, where

  15. Burn-up characteristics of ADS system utilizing the fuel composition from MOX PWRs spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marsodi E-mail: marsodi@batan.go.id; Lasman, K.A.S.; Nishihara, K. E-mail: nishi@omega.tokai.jaeri.go.jp; Osugi, T.; Tsujimoto, K.; Marsongkohadi; Su' ud, Z. E-mail: szaki@fi.itb.ac.id

    2002-12-01

    Burn-up characteristics of accelerator-driven system, ADS has been evaluated utilizing the fuel composition from MOX PWRs spent fuel. The system consists of a high intensity proton beam accelerator, spallation target, and sub-critical reactor core. The liquid lead-bismuth, Pb-Bi, as spallation target, was put in the center of the core region. The general approach was conducted throughout the nitride fuel that allows the utilities to choose the strategy for destroying or minimizing the most dangerous high level wastes in a fast neutron spectrum. The fuel introduced surrounding the target region was the same with the composition of MOX from 33 GWd/t PWRs spent-fuel with 5 year cooling and has been compared with the fuel composition from 45 and 60 GWd/t PWRs spent-fuel with the same cooling time. The basic characteristics of the system such as burn-up reactivity swing, power density, neutron fluxes distribution, and nuclides densities were obtained from the results of the neutronics and burn-up analyses using ATRAS computer code of the Japan Atomic Energy research Institute, JAERI.

  16. GLOBAL DECOUPLING ON THE RHIC RAMP.

    Energy Technology Data Exchange (ETDEWEB)

    LUO, Y.; CAMERON, P.; DELLA PENNA, A.; FISCHER, W.; ET AL.

    2005-05-16

    The global betatron decoupling on the ramp is an important issue for the operation of the Relativistic Heavy Ion Collider (RHIC), especially in the RHIC polarized proton (pp) run. To avoid the major betatron and spin resonances on the ramp, the betatron tunes are constrained. And the rms value of the vertical closed orbit should be smaller than 0.5mm. Both require the global coupling on the ramp to be well corrected. Several ramp decoupling schemes were found and tested at RHIC, like N-turn map decoupling, three-ramp correction, coupling amplitude modulation, and coupling phase modulation. In this article, the principles of these methods are shortly reviewed and compared. Among them, coupling angle modulation is a robust and fast one. It has been applied to the global decoupling in the routine RHIC operation.

  17. Size Design of CdZnTe Detector Shield for Measuring Burnup of Spent Fuel

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    <正>It is important to measure the burnup of spent fuel for nuclear safeguards, burnup credit and critical safety in spent-fuel reprocessing process. The purpose of this work is designing a portable device to

  18. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  19. PWR fuel performance and burnup extension in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yokote, M. [Kansai Electric Power Co., Inc., Osaka (Japan); Kondo, Y.; Abeta, S.

    1996-10-01

    Japanese utilities and fuel manufacturers have expanded much of their resources and efforts to maintain a reliable supply of PWR fuel for Japan. In the early 1970s, since the level of knowledge and experience of using fuel was less than now, some problems were encountered. However, their causes were investigated and countermeasures implemented, the design improved and quality control enhanced. The results can already be seen by significantly improved performance of the PWR plants now in operation, frequency of problems was quickly reduced. Since fuel reliability has been improved, the emphasis has shifted to improving economics by increasing burnup and using uranium resources effectively. The maximum discharged burnup was previously limited to 39 GWd/t and STEP1 burnup extension to 48 GWd/t has been gradually developed, while STEP2 burnup extension to 55 GWd/t is started to be demonstrated from 1996. Because resources in Japan are scarce, a policy was selected of conserving and making effective use of these resources by recycling the uranium and plutonium recovered from reactors. Consequently, significant work is being done on the development of MOX fuel and utilization of recovered uranium. (author)

  20. Need for higher fuel burnup at the Hatch Plant

    Energy Technology Data Exchange (ETDEWEB)

    Beckhman, J.T. [Georgia Power Co., Birmingham, AL (United States)

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make the conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.

  1. Lean Application to Manufacturing Ramp-Up

    DEFF Research Database (Denmark)

    Christensen, Irene; Rymaszewska, Anna

    2016-01-01

    product development process, that is, the ramp-up process, is a critical, early enabler of lean manufacturing. The manufacturing strategy literature conceptualizes a state of “leanness in operations,” which can consolidate both the concepts of lean and manufacturing ramp-up, providing a dual perspective......This article provides a theoretical overview of the concepts of lean and manufacturing ramp-up in an attempt to conceptualize the strategic areas in which lean philosophy and principles can be applied for continuous improvements. The application of lean principles during the final stage of a new....... Abstracting from the extant literature, the authors considered the competitiveness of manufacturing companies from two principal perspectives: the leanness of the ramp-up process and the new-value creation of quality managers. While much of the literature fails to acknowledge that the roots of lean actually...

  2. Wind-Friendly Flexible Ramping Product Design in Multi-Timescale Power System Operations

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Mingjian; Zhang, Jie; Wu, Hongyu; Hodge, Bri-Mathias

    2017-07-01

    With increasing wind power penetration in the electricity grid, system operators are recognizing the need for additional flexibility, and some are implementing new ramping products as a type of ancillary service. However, wind is generally thought of as causing the need for ramping services, not as being a potential source for the service. In this paper, a multi-timescale unit commitment and economic dispatch model is developed to consider the wind power ramping product (WPRP). An optimized swinging door algorithm with dynamic programming is applied to identify and forecast wind power ramps (WPRs). Designed as positive characteristics of WPRs, the WPRP is then integrated into the multi-timescale dispatch model that considers new objective functions, ramping capacity limits, active power limits, and flexible ramping requirements. Numerical simulations on the modified IEEE 118-bus system show the potential effectiveness of WPRP in increasing the economic efficiency of power system operations with high levels of wind power penetration. It is found that WPRP not only reduces the production cost by using less ramping reserves scheduled by conventional generators, but also possibly enhances the reliability of power system operations. Moreover, wind power forecasts play an important role in providing high-quality WPRP service.

  3. 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.

  4. Design and construction of a prototype advanced on-line fuel burn-up monitoring system for the modular pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Su, Bingjing; Hawari, Ayman, I.

    2004-03-30

    Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor currently under study as a next generation reactor system. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burn-up limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burn-up limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated approaches to analyzing fuel pebbles in real time using gamma spectroscopy and possibly using passive neutron counting of spontaneous fission neutrons to provide the speed, accuracy, and burn-up range required for burnup determination of MPBR. It involved all phases necessary to develop and construct a burn-up monitor, including a review of the design requirements of the system, identification of detection methodologies, modeling and development of potential designs, and finally, the construction and testing of an operational detector system. Based upon the research work performed in this project, the following conclusions are made. In terms of using gamma spectrometry, two possible approaches were identified for burnup assay. The first approach is based on the measurement of the absolute activity of Cs-137. However, due to spectral interference and the need for absolute calibration of the spectrometer, the uncertainty in burnup determination using this approach was found to range from {approx} {+-}40% at beginning of life to {approx} {+-}10% at the discharge burnup. An alternative approach is to use a relative burnup indicator. In this

  5. Optimizing density down-ramp injection for beam-driven plasma wakefield accelerators

    Science.gov (United States)

    Martinez de la Ossa, A.; Hu, Z.; Streeter, M. J. V.; Mehrling, T. J.; Kononenko, O.; Sheeran, B.; Osterhoff, J.

    2017-09-01

    Density down-ramp (DDR) injection is a promising concept in beam-driven plasma wakefield accelerators for the generation of high-quality witness beams. We review and complement the theoretical principles of the method and employ particle-in-cell (PIC) simulations in order to determine constrains on the geometry of the density ramp and the current of the drive beam, regarding the applicability of DDR injection. Furthermore, PIC simulations are utilized to find optimized conditions for the production of high-quality beams. We find and explain the intriguing result that the injection of an increased charge by means of a steepened ramp favors the generation of beams with lower emittance. Exploiting this fact enables the production of beams with high charge (˜140 pC ), low normalized emittance (˜200 nm ) and low uncorrelated energy spread (0.3%) in sufficiently steep ramps even for drive beams with moderate peak current (˜2.5 kA ).

  6. Accelerating Science Driven System Design With RAMP

    Energy Technology Data Exchange (ETDEWEB)

    Wawrzynek, John [Univ. of California, Berkeley, CA (United States)

    2015-05-01

    Researchers from UC Berkeley, in collaboration with the Lawrence Berkeley National Lab, are engaged in developing an Infrastructure for Synthesis with Integrated Simulation (ISIS). The ISIS Project was a cooperative effort for “application-driven hardware design” that engages application scientists in the early parts of the hardware design process for future generation supercomputing systems. This project served to foster development of computing systems that are better tuned to the application requirements of demanding scientific applications and result in more cost-effective and efficient HPC system designs. In order to overcome long conventional design-cycle times, we leveraged reconfigurable devices to aid in the design of high-efficiency systems, including conventional multi- and many-core systems. The resulting system emulation/prototyping environment, in conjunction with the appropriate intermediate abstractions, provided both a convenient user programming experience and retained flexibility, and thus efficiency, of a reconfigurable platform. We initially targeted the Berkeley RAMP system (Research Accelerator for Multiple Processors) as that hardware emulation environment to facilitate and ultimately accelerate the iterative process of science-driven system design. Our goal was to develop and demonstrate a design methodology for domain-optimized computer system architectures. The tangible outcome is a methodology and tools for rapid prototyping and design-space exploration, leading to highly optimized and efficient HPC systems.

  7. Aphotic zone carbonate production on a Miocene ramp, Central Apennines, Italy

    Science.gov (United States)

    Corda, Laura; Brandano, Marco

    2003-09-01

    The lower Miocene Latium-Abruzzi platform was a low-angle ramp that developed under tropical-to-subtropical conditions, but was dominated by bryomol and rhodalgal sediment associations. The Aquitanian to Serravallian sequence described here paraconformably overlies the Cretaceous limestones. It consists of a lowstand systems tract, a transgressive systems tract and a highstand systems tract. Based on facies analysis and on the light dependence of biotic associations, the ramp is divided into three parts: an inner ramp, a middle ramp and an outer ramp. The inner ramp facies are represented by a few metres of coral framestone, rhodolith floatstone-rudstone and balanid macroids floatstone without wave-related structures. The middle ramp consists of structureless bioclastic grainstone to packstone, floatstone and rudstone with rhodoliths and larger foraminifera. The outer ramp facies—proximal sector—are composed of crudely stratified bryozoan-dominated packstone to floatstone which extend over the whole platform. The outer ramp facies—intermediate sector—are represented by wackestone, packstone and rarely grainstone with foraminifera and echinoid fragments. The final depositional profile of the ramp was strongly influenced by the main organisms producing sediment. During the lowstand, the resulting profile is a ramp type. During the transgressive phase, the rapid spreading of the outer ramp facies belt, as a consequence of the enhanced productivity of the light-independent biota, is believed to be promoted by a change from oligotrophic to eutrophic conditions. Climate and/or tectonics are presumed to have played an important role in continental runoff and then in the nutrients delivery. During the highstand phase, the system returns to rates of production uniform throughout the platform. The high rates of carbonate production occurring in the aphotic zone are quite unusual in tropical settings and represent a provocative trend in apparent contrast with the

  8. OREST - The hammer-origen burnup program system

    Energy Technology Data Exchange (ETDEWEB)

    Hesse, U. (Gesellschaft fur Reaktorsicherheit mbH Forschungsgelande, 8046 Garching bei Munchen (DE))

    1988-08-01

    Reliable prediction of the characteristics of irradiated light water reactor fuels (e.g., afterheat power, neutron and gamma radiation sources, final uranium and plutonium contents) is needed for many aspects of the nuclear fuel cycle. Two main problems must be solved: the simulation of all isotopic nuclear reactions and the simulation of neutron fluxes setting the reactions in motion. In state-of-the-art computer techniques, a combination of specialized codes for lattice cell and burnup calculations is preferred to solve these cross-linked problems in time or burnup step approximation. In the program system OREST, developed for official and commercial tasks in the Federal Republic of Germany nuclear fuel cycle, the well-known codes HAMMER and ORIGEN and directly coupled with a fuel rod temperature module.

  9. Off-ramps and on-ramps: keeping talented women on the road to success.

    Science.gov (United States)

    Hewlett, Sylvia Ann; Luce, Carolyn Buck

    2005-03-01

    Most professional women step off the career fast track at some point. With children to raise, elderly parents to care for, and other pulls on their time, these women are confronted with one off-ramp after another. When they feel pushed at the same time by long hours and unsatisfying work, the decision to leave becomes even easier. But woe to the woman who intends for that exit to be temporary. The on-ramps for professional women to get back on track are few and far between, the authors confirm. Their new survey research reveals for the first time the extent of the problem--what percentage of highly qualified women leave work and for how long, what obstacles they face coming back, and what price they pay for their time-outs. And what are the implications for corporate America? One thing at least seems clear: As market and economic factors align in ways guaranteed to make talent constraints and skill shortages huge issues again, employers must learn to reverse this brain drain. Like it or not, large numbers of highly qualified, committed women need to take time out of the workplace. The trick is to help them maintain connections that will allow them to reenter the workforce without being marginalized for the rest of their lives. Strategies for building such connections include creating reduced-hour jobs, providing flexibility in the workday and in the arc of a career, removing the stigma of taking time off, refusing to burn bridges, offering outlets for altruism, and nurturing women's ambition.

  10. 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)

  11. Fuel burnup calculation of a research reactor plate element

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: nadiasam@gmail.com, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work consists in simulating the burnup of two different plate type fuel elements, where one is the benchmark MTR of the IAEA, which is made of an alloy of uranium and aluminum, while the other belonging to a typical multipurpose reactor is composed of an alloy of uranium and silicon. The simulation is performed using the WIMSD-5B computer code, which makes use of deterministic methods for solving neutron transport. In developing this task, fuel element equivalent cells were calculated representing each of the reactors to obtain the initial concentrations of each isotope constituent element of the fuel cell and the thicknesses corresponding to each region of the cell, since this information is part of the input data. The compared values of the k∞ showed a similar behavior for the case of the MTR calculated with the WIMSD-5B and EPRI-CELL codes. Relating the graphs of the concentrations in the burnup of both reactors, there are aspects very similar to each isotope selected. The application WIMSD-5B code to calculate isotopic concentrations and burnup of the fuel element, proved to be satisfactory for the fulfillment of the objective of this work. (author)

  12. Microprocessor-controlled, programmable ramp voltage generator

    Energy Technology Data Exchange (ETDEWEB)

    Hopwood, J.

    1978-11-01

    A special-purpose voltage generator has been developed for driving the quadrupole mass filter of a residual gas analyzer. The generator is microprocessor-controlled with desired ramping parameters programmed by setting front-panel digital thumb switches. The start voltage, stop voltage, and time of each excursion are selectable. A maximum of five start-stop levels may be pre-selected for each program. The ramp voltage is 0 to 10 volts with sweep times from 0.1 to 999.99 seconds.

  13. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  14. Modeling of burnup express-estimation for UO{sub 2}-fuel

    Energy Technology Data Exchange (ETDEWEB)

    Likhanskii, Vladimir V.; Tokarev, Sergey A.; Vilkhivskaya, Olga V., E-mail: vilhivskaya_olga@mail.ru

    2017-03-15

    Highlights: • Proposed engineering model estimates fuel burnup by {sup 134}Cs/{sup 137}Cs activity ratio. • Buildup of cesium isotopes relies on changing neutron spectrum in the core cycle. • {sup 134}Cs/{sup 137}Cs activity ratios in FAs with Gd-doped fuel rods are analyzed. • Comparison of the model calculations with the NPPs spike measurements is presented. - Abstract: The paper presents the developed engineering model of cesium isotopes production as function of UO{sub 2}-fuel burnup and an assessment of their activity ratios. The model considers the evolution of linear power of gadolinium-doped fuel rods and fuel rods surrounding them in fuel assemblies with high enrichment fuel, harder neutron spectrum, and the changes in cross-sections of neutron reactions in thermal and epithermal energy areas. Parametrical dependences in the model are based on the fuel operation data for nuclear power plants and on the detailed neutronic-physical calculations of the core. Presented are the results of the model calculations for the {sup 134}Cs/{sup 137}Cs activity ratios in fuel taking into account the parameter of hardness of the neutron spectrum during the first irradiation cycle for fuel with enrichment ranging from 3.6 wt% in {sup 235}U.

  15. Risk assessment in ramps for heavy vehicles--A French study.

    Science.gov (United States)

    Cerezo, Veronique; Conche, Florence

    2016-06-01

    This paper presents the results of a study dealing with the risk for heavy vehicles in ramps. Two approaches are used. On one hand, statistics are applied on several accidents databases to detect if ramps are more risky for heavy vehicles and to define a critical value for longitudinal slope. χ(2) test confirmed the risk in ramps and statistical analysis proved that a longitudinal slope superior to 3.2% represents a higher risk for heavy vehicles. On another hand, numerical simulations allow defining the speed profile in ramps for two types of heavy vehicles (tractor semi-trailer and 2-axles rigid body) and different loads. The simulations showed that heavy vehicles must drive more than 1000 m on ramps to reach their minimum speed. Moreover, when the slope is superior to 3.2%, tractor semi-trailer presents a strong decrease of their speed until 50 km/h. This situation represents a high risk of collision with other road users which drive at 80-90 km/h. Thus, both methods led to the determination of a risky configuration for heavy vehicles: ramps with a length superior to 1000 m and a slope superior to 3.2%. An application of this research work concerns design methods and guidelines. Indeed, this study provides threshold values than can be used by engineers to make mandatory specific planning like a lane for slow vehicles.

  16. Double Ramp Loss Based Reject Option Classifier

    Science.gov (United States)

    2015-05-22

    Cao et al. (Eds.): PAKDD 2015, Part I, LNAI 9077, pp. 151–163, 2015. DOI: 10.1007/978-3-319-18038-0 12 152 N. Manwani et al. ρ is the parameter...A.: kernlab - an S4 package for kernel methods in R. Journal of Statistical Software 11(9), 1–20 ( 2004 ) Double Ramp Loss Based Reject Option

  17. Tension Recovery following Ramp-Shaped Release in High-Ca and Low-Ca Rigor Muscle Fibers: Evidence for the Dynamic State of AMADP Myosin Heads in the Absence of ATP

    Science.gov (United States)

    Sugi, Haruo; Yamaguchi, Maki; Ohno, Tetsuo; Kobayashi, Takakazu; Chaen, Shigeru; Okuyama, Hiroshi

    2016-01-01

    During muscle contraction, myosin heads (M) bound to actin (A) perform power stroke associated with reaction, AMADPPi → AM + ADP + Pi. In this scheme, A • M is believed to be a high-affinity complex after removal of ATP. Biochemical studies on extracted protein samples show that, in the AM complex, actin-binding sites are located at both sides of junctional peptide between 50K and 20K segments of myosin heavy chain. Recently, we found that a monoclonal antibody (IgG) to the junctional peptide had no effect on both in vitro actin-myosin sliding and skinned muscle fiber contraction, though it covers the actin-binding sites on myosin. It follows from this that, during muscle contraction, myosin heads do not pass through the static rigor AM configuration, determined biochemically and electron microscopically using extracted protein samples. To study the nature of AM and AMADP myosin heads, actually existing in muscle, we examined mechanical responses to ramp-shaped releases (0.5% of Lo, complete in 5ms) in single skinned rabbit psoas muscle fibers in high-Ca (pCa, 4) and low-Ca (pCa, >9) rigor states. The fibers exhibited initial elastic tension drop and subsequent small but definite tension recovery to a steady level. The tension recovery was present over many minutes in high-Ca rigor fibers, while it tended to decrease quickly in low-Ca rigor fibers. EDTA (10mM, with MgCl2 removed) had no appreciable effect on the tension recovery in high-Ca rigor fibers, while it completely eliminated the tension recovery in low-Ca rigor fibers. These results suggest that the AMADP myosin heads in rigor muscle have long lifetimes and dynamic properties, which show up as the tension recovery following applied release. Possible AM linkage structure in muscle is discussed in connection with the X-ray diffraction pattern from contracting muscle, which is intermediate between resting and rigor muscles. PMID:27583360

  18. Tension Recovery following Ramp-Shaped Release in High-Ca and Low-Ca Rigor Muscle Fibers: Evidence for the Dynamic State of AMADP Myosin Heads in the Absence of ATP.

    Science.gov (United States)

    Sugi, Haruo; Yamaguchi, Maki; Ohno, Tetsuo; Kobayashi, Takakazu; Chaen, Shigeru; Okuyama, Hiroshi

    2016-01-01

    During muscle contraction, myosin heads (M) bound to actin (A) perform power stroke associated with reaction, AMADPPi → AM + ADP + Pi. In this scheme, A • M is believed to be a high-affinity complex after removal of ATP. Biochemical studies on extracted protein samples show that, in the AM complex, actin-binding sites are located at both sides of junctional peptide between 50K and 20K segments of myosin heavy chain. Recently, we found that a monoclonal antibody (IgG) to the junctional peptide had no effect on both in vitro actin-myosin sliding and skinned muscle fiber contraction, though it covers the actin-binding sites on myosin. It follows from this that, during muscle contraction, myosin heads do not pass through the static rigor AM configuration, determined biochemically and electron microscopically using extracted protein samples. To study the nature of AM and AMADP myosin heads, actually existing in muscle, we examined mechanical responses to ramp-shaped releases (0.5% of Lo, complete in 5ms) in single skinned rabbit psoas muscle fibers in high-Ca (pCa, 4) and low-Ca (pCa, >9) rigor states. The fibers exhibited initial elastic tension drop and subsequent small but definite tension recovery to a steady level. The tension recovery was present over many minutes in high-Ca rigor fibers, while it tended to decrease quickly in low-Ca rigor fibers. EDTA (10mM, with MgCl2 removed) had no appreciable effect on the tension recovery in high-Ca rigor fibers, while it completely eliminated the tension recovery in low-Ca rigor fibers. These results suggest that the AMADP myosin heads in rigor muscle have long lifetimes and dynamic properties, which show up as the tension recovery following applied release. Possible AM linkage structure in muscle is discussed in connection with the X-ray diffraction pattern from contracting muscle, which is intermediate between resting and rigor muscles.

  19. Road and Street Centerlines - RAMPS_INDOT_IN: Ramp System in Indiana (Indiana Department of Transportation, Line Shapefile)

    Data.gov (United States)

    NSGIC GIS Inventory (aka Ramona) — RAMPS_INDOT_IN is a line shapefile that contains all ramps in Indiana, provided by personnel of Indiana Department of Transportation (INDOT), Business Information...

  20. Anatomy of a cyclically packaged Mesoproterozoic carbonate ramp in northern Canada

    Science.gov (United States)

    Sherman, A. G.; Narbonne, G. M.; James, N. P.

    2001-03-01

    Carbonates in the upper member of the Mesoproterozoic Victor Bay Formation are dominated by lime mud and packaged in cycles of 20-50 m. These thicknesses exceed those of classic shallowing-upward cycles by almost a factor of 10. Stratigraphic and sedimentological evidence suggests high-amplitude, high-frequency glacio-eustatic cyclicity, and thus a cool global climate ca. 1.2 Ga. The Victor Bay ramp is one of several late Proterozoic carbonate platforms where the proportions of lime mud, carbonate grains, and microbialites are more typical of younger Phanerozoic successions which followed the global waning of stromatolites. Facies distribution in the study area is compatible with deposition on a low-energy, microtidal, distally steepened ramp. Outer-ramp facies are hemipelagic lime mudstone, shale, carbonaceous rhythmite, and debrites. Mid-ramp facies are molar-tooth limestone tempestite with microspar-intraclast lags. In a marine environment where stromatolitic and oolitic facies were otherwise rare, large stromatolitic reefs developed at the mid-ramp, coeval with inner-ramp facies of microspar grainstone, intertidal dolomitic microbial laminite, and supratidal evaporitic red shale. Deep-subtidal, outer-ramp cycles occur in the southwestern part of the study area. Black dolomitic shale at the base is overlain by ribbon, nodular, and carbonaceous carbonate facies, all of which exhibit signs of synsedimentary disruption. Cycles in the northeast are shallow-subtidal and peritidal in character. Shallow-subtidal cycles consist of basal deep-water facies, and an upper layer of subtidal molar-tooth limestone tempestite interbedded with microspar calcarenite facies. Peritidal cycles are identical to shallow-subtidal cycles except that they contain a cap of dolomitic tidal-flat microbial laminite, and rarely of red shale sabkha facies or of sandy polymictic conglomerate. A transect along the wall of a valley extending 8.5 km perpendicular to depositional strike reveals

  1. 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

  2. Evaluation of Isotopic Measurements and Burn-up Value of Sample GU3 of ARIANE Project

    Energy Technology Data Exchange (ETDEWEB)

    Tore, C.; Rodriguez Rivada, A.

    2014-07-01

    Estimation of the burn-up value of irradiated fuel and its isotopic composition are important for criticality analysis, spent fuel management and source term estimation. The practical way to estimate the irradiated fuel composition and burn.up value is calculation with validated code and nuclear data. Such validation of the neutronic codes and nuclear data requires the benchmarking with measured values. (Author)

  3. X-1E Loaded in B-29 Mothership on Ramp

    Science.gov (United States)

    1955-01-01

    The Bell Aircraft Corporation X-1E airplane being loaded under the mothership, Boeing B-29. The X-planes had originally been lowered into a loading pit and the launch aircraft towed over the pit, where the rocket plane was hoisted by belly straps into the bomb bay. By the early 1950s a hydraulic lift had been installed on the ramp at the NACA High-Speed Flight Station to elevate the launch aircraft and then lower it over the rocket plane for mating.

  4. Forward modeling of shock-ramped tantalum

    Science.gov (United States)

    Brown, Justin L.; Carpenter, John H.; Seagle, Christopher T.

    2017-01-01

    Dynamic materials experiments on the Z-machine are beginning to reach a regime where traditional analysis techniques break down. Time dependent phenomena such as strength and phase transition kinetics often make the data obtained in these experiments difficult to interpret. We present an inverse analysis methodology to infer the equation of state (EOS) from velocimetry data in these types of experiments, building on recent advances in the propagation of uncertain EOS information through a hydrocode simulation. An example is given for a shock-ramp experiment in which tantalum was shock compressed to 40 GPa followed by a ramp to 80 GPa. The results are found to be consistent with isothermal compression and Hugoniot data in this regime.

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

    Directory of Open Access Journals (Sweden)

    Iqbal Masood

    2008-01-01

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

  6. Research on recognition of ramp angle based on transducer

    Directory of Open Access Journals (Sweden)

    Wenhao GU

    2015-12-01

    Full Text Available Focusing on the recognition of ramp angle, the relationship between the signal of vehicle transducer and real ramp angle is studied. The force change of vehicle on the ramp, and the relationship between the body tilt angle and front and rear suspension scale is discussed. According to the suspension and tire deformation, error angle of the ramp angle is deduced. A mathematical model is established with Matlab/Simulink and used for simulation to generate error curve of ramp angle. The results show that the error angle increases with the increasing of the ramp angle, and the limit value can reach 6.5%, while the identification method can effectively eliminate this error, and enhance the accuracy of ramp angle recognition.

  7. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

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

    Science.gov (United States)

    Lim, Harn Chyi; Rudman, Karin; Krishnan, Kapil; McDonald, Robert; Dickerson, Patricia; Gong, Bowen; Peralta, Pedro

    2016-08-01

    Diffusion of fission gases in UO2 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 UO2 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.

  9. Preparation of data relevant to ''Equivalent Uniform Burnup'' and Equivalent Initial Enrichment'' for burnup credit evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan)

    2001-11-01

    Based on the PWR spent fuel composition data measured at JAERI, two kinds of simplified methods such as ''Equivalent Uniform Burnup'' and ''Equivalent Initial Enrichment'' have been introduced. And relevant evaluation curves have been prepared for criticality safety evaluation of spent fuel storage pool and transport casks, taking burnup of spent fuel into consideration. These simplified methods can be used to obtain an effective neutron multiplication factor for a spent fuel storage/transportation system by using the ORIGEN2.1 burnup code and the KENO-Va criticality code without considering axial burnup profile in spent fuel and other various factors introducing calculated errors. ''Equivalent Uniform Burnup'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis, in which the experimentally obtained isotopic composition together with a typical axial burnup profile and various factors such as irradiation history are considered on the conservative side. On the other hand, Equivalent Initial Enrichment'' is set up for its criticality analysis to be reactivity equivalent with the detailed analysis such as above when it is used in the so called fresh fuel assumption. (author)

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

    Directory of Open Access Journals (Sweden)

    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

  11. 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

  12. X-ray diffraction of molybdenum under ramp compression to 1 TPa

    Science.gov (United States)

    Wang, Jue; Coppari, Federica; Smith, Raymond F.; Eggert, Jon H.; Lazicki, Amy E.; Fratanduono, Dayne E.; Rygg, J. Ryan; Boehly, Thomas R.; Collins, Gilbert W.; Duffy, Thomas S.

    2016-09-01

    Molybdenum (Mo) is a transition metal with a wide range of technical applications. There has long been strong interest in its high-pressure behavior, and it is often used as standard for high-pressure experiments. Combining powder x-ray diffraction and dynamic ramp compression, structural and equation of state data were collected for solid Mo to 1 TPa (10 Mbar). Diffraction results are consistent with Mo remaining in the body-centered-cubic structure into the TPa regime. Stress-density data show that Mo under ramp loading is less compressible than the room-temperature isotherm but more compressible than the single-shock Hugoniot.

  13. Evaluation of genetic diversity of Clinacanthus nutans (Acanthaceaea) using RAPD, ISSR and RAMP markers.

    Science.gov (United States)

    Ismail, Noor Zafirah; Arsad, Hasni; Samian, Mohammed Razip; Ab Majid, Abdul Hafiz; Hamdan, Mohammad Razak

    2016-10-01

    Three polymerase chain reaction (PCR) techniques were compared to analyse the genetic diversity of Clinacanthus nutans eight populations in the northern region of Peninsular Malaysia. The PCR techniques were random amplified polymorphic deoxyribonucleic acids (RAPD), inter-simple sequence repeats (ISSR) and random amplified microsatellite polymorphisms (RAMP). Leaf genomic DNA was PCR amplified using 17 RAPD, 8 ISSR and 136 RAMP primers . However, only 10 RAPD primers, 5 ISSR primers and 37 RAMP primers produced reproducible bands. The results were evaluated for polymorphic information content (PIC), marker index (MI) and resolving power (RP). The RAMP marker was the most useful marker compared to RAPD and ISSR markers because it showed the highest average value of PIC (0.25), MI (11.36) and RP (2.86). The genetic diversity showed a high percentage of polymorphism at the species level compared to the population level. Furthermore, analysis of molecular variance revealed that the genetic diversity was higher within populations, as compared to among populations of C. nutans. From the results, the RAMP technique was recommended for the analysis of genetic diversity of C. nutans.

  14. Assigning on-ramp flows to maximize capacity of highway with two on-ramps and one off-ramp in between

    Science.gov (United States)

    Chen, Jing; Lin, Lan; Jiang, Rui

    2017-01-01

    In this paper, we study the capacity of a highway with two on-ramps and one off-ramp in between by using a cellular automaton traffic flow model. We investigate how to maximize the system capacity by assigning proper traffic flow to the two on-ramps. The system phase diagram is presented and eight different regions are observed under different conditions. It is shown that in region I, in which both on-ramps are in free flow and the main road upstream of the upstream on-ramp is in congestion, assigning proper proportion of the demand to two on-ramps could maximize the system capacity. Two critical values of the off-ramp flow ratio poff have been observed. When poff p off , c 2, no demand should be assigned to the upstream on-ramp. An analytical investigation has been performed to calculate the critical values. The analytical results are in good agreement with the simulation ones.

  15. Fully Coupled Modeling of Burnup-Dependent (U1- y , Pu y )O2- x Mixed Oxide Fast Reactor Fuel Performance

    Science.gov (United States)

    Liu, Rong; Zhou, Wenzhong; Zhou, Wei

    2016-03-01

    During the fast reactor nuclear fuel fission reaction, fission gases accumulate and form pores with the increase of fuel burnup, which decreases the fuel thermal conductivity, leading to overheating of the fuel element. The diffusion of plutonium and oxygen with high temperature gradient is also one of the important fuel performance concerns as it will affect the fuel material properties, power distribution, and overall performance of the fuel pin. In order to investigate these important issues, the (U1- y Pu y )O2- x fuel pellet is studied by fully coupling thermal transport, deformation, oxygen diffusion, fission gas release and swelling, and plutonium redistribution to evaluate the effects on each other with burnup-dependent models, accounting for the evolution of fuel porosity. The approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with fast reactor mixed oxide fuel performance analysis. The modeling results showed a consistent fuel performance comparable with the previous results. Burnup degrades the fuel thermal conductivity, resulting in a significant fuel temperature increase. The fission gas release increased rapidly first and then steadily with the burnup increase. The fuel porosity increased dramatically at the beginning of the burnup and then kept constant as the fission gas released to the fuel free volume, causing the fuel temperature to increase. Another important finding is that the deviation from stoichiometry of oxygen affects greatly not only the fuel properties, for example, thermal conductivity, but also the fuel performance, for example, temperature distribution, porosity evolution, grain size growth, fission gas release, deformation, and plutonium redistribution. Special attention needs to be paid to the deviation from stoichiometry of oxygen in fuel fabrication. Plutonium content will also affect the fuel material properties and performance

  16. The high burn-up structure in nuclear fuel

    Directory of Open Access Journals (Sweden)

    Vincenzo V. Rondinella

    2010-12-01

    Full Text Available During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today's light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation.

  17. 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

  18. 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.

  19. Burnup calculations for the HOMER-15 and SAFE-300 reactors

    Science.gov (United States)

    Amiri, Benjamin W.; Poston, David I.

    2002-01-01

    The Heatpipe Power System (HPS) is a near-term low-cost space fission power system. As the U-235 fuel of the HPS is burned, higher actinides and fission products will be produced. This will cause changes in system reactivity, radioactivity, and decay power. One potential concern is that gaseous fission products may exert excessive pressure on the fuel pin cladding. To evaluate these issues, simulations were run in MONTEBURNS. MONTEBURNS is an automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. This paper describes the results of these simulations, as well as how those results compare with the current experimental database of irradiated materials. .

  20. 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.

  1. 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.

  2. Response of YBCO/PCBO/YBCO ramp type Josephson junctions to near MM wave irradiation

    NARCIS (Netherlands)

    Terpstra, D.; Gerritsma, G.J.; Rogalla, H.; Gupta, R.; Hu, Q.Y.

    1994-01-01

    A high Tc Josephson device for high frequency detection applications is being developed, consisting of an YBCO/PBCO/YBCO ramp type junction and a broad band log-periodic antenna. In this contribution we present the response of such a device to (near) mm wave irradiation. Shapiro steps have been obse

  3. Numerical solution of shock and ramp compression for general material properties

    Energy Technology Data Exchange (ETDEWEB)

    Swift, D C

    2009-01-28

    A general formulation was developed to represent material models for applications in dynamic loading. Numerical methods were devised to calculate response to shock and ramp compression, and ramp decompression, generalizing previous solutions for scalar equations of state. The numerical methods were found to be flexible and robust, and matched analytic results to a high accuracy. The basic ramp and shock solution methods were coupled to solve for composite deformation paths, such as shock-induced impacts, and shock interactions with a planar interface between different materials. These calculations capture much of the physics of typical material dynamics experiments, without requiring spatially-resolving simulations. Example calculations were made of loading histories in metals, illustrating the effects of plastic work on the temperatures induced in quasi-isentropic and shock-release experiments, and the effect of a phase transition.

  4. Paths to ignition by radio frequency heating during the B-field ramp

    Science.gov (United States)

    Myra, J. R.; Aamodt, R. E.; D'Ippolito, D. A.

    2000-05-01

    To conserve transformer volt-seconds, power to toroidal magnetic field coils, and to trigger an early transition into high confinement (H) mode, where the requirements on auxiliary power are lower, rf heating during the B-field ramp phase of ignition-class tokamaks is considered. The scheme is analyzed by modifying the usual plasma operating condition diagrams to apply to the ramp phase where the magnetic field, plasma current, and density are changing. It is shown that ion cyclotron range-of-frequencies direct electron heating during the ramp phase of IGNITOR [B. Coppi, M. Nassi, and L. E. Sugiyama, Phys. Scr. 45, 112 (1992)], as proposed by Majeski [R. Majeski, in AIP Conference Proceedings 485—Radio Frequency Power in Plasmas, Annapolis, MD (AIP, New York, 1999), p. 353], may be useful in optimizing the operating condition path to ignition.

  5. Validation of Up-the-Ramp Sampling with Cosmic Ray Rejection on IR Detectors

    CERN Document Server

    Offenberg, J D; Rauscher, B J; Forrest, W J; Hanisch, R J; Mather, J C; McKelvey, M E; McMurray, R E; Nieto-Santisteban, M A; Pipher, J L; Gupta, R; Stockman, H S

    2000-01-01

    We examine cosmic ray rejection methodology on data collected from InSb and Si:As detectors. The application of an Up-the-Ramp sampling technique with on-the-fly cosmic ray identification and mitigation is the focus of this study. This technique is valuable for space-based observatories which are exposed to high-radiation environments. We validate the Up-the-Ramp approach on radiation-test data sets with InSb and Si:As detectors which were generated for SIRTF. The Up-the-Ramp sampling method studied in this paper is over 99.9% effective at removing cosmic rays and preserves the structure and photometric quality of the image to well within the measurement error.

  6. 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)

  7. Evaluatie ramp in Bhopal in relatie tot arbeidsveiligheid

    OpenAIRE

    1987-01-01

    De ramp in Bhopal wordt geanalyseerd op het punt van procesveiligheid en arbeidsveiligheid. Geconcludeerd wordt dat de ramp te wijten is aan een samenloop van omstandigheden, een oorzakenstructuur. Met behulp van de TNO ongevallen - databank FACTS is geconcludeerd dat deze structuur nauwelijks afwijkt van andere ernstige ongevallen. De ramp in Bhopal was dan ook uitsluitend een uniek ongeval door de zeer fatale gevolgen. De belangrijkste lessen die uit Bhopal geleerd kunnen worden zijn ; het ...

  8. Entry ramps in the Nagel-Schreckenberg model

    DEFF Research Database (Denmark)

    Pedersen, Morten Monrad; Ruhoff, Peder Thusgaard

    2002-01-01

    This paper describes a way of including entry ramps in the Nagel-Schreckenberg traffic model. The idea is to place what are called shadow cars on a highway next to cars on entry ramps, which enables the drivers to take ramp cars into account. The model is shown to capture important real......-life traffic phenomena that have not been included in previous models. Furthermore, it is demonstrated that the desirable properties of the Nagel-Schreckenberg model are retained....

  9. 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

  10. Research on Power Ramp Testing Method for PWR Fuel Rod at Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In order to develop high performance fuel assembly for domestic nuclear power plant, it is necessary to master some fundamental test technology. So the research on the power ramp testing methods is proposed. A tentative power ramp test for short PWR fuel rod has been conducted at the heavy water research reactor (HWRR) in China Institute of Atomic Energy (CIAE) in May of 2001. The in-pile test rig was placed into the central channel of the reactor . The test rig consists of pressure pipe assembly, thimble, solid neutron absorbing screen and its driving parts, etc.. The test

  11. Ramp compression of diamond to five terapascals.

    Science.gov (United States)

    Smith, R F; Eggert, J H; Jeanloz, R; Duffy, T S; Braun, D G; Patterson, J R; Rudd, R E; Biener, J; Lazicki, A E; Hamza, A V; Wang, J; Braun, T; Benedict, L X; Celliers, P M; Collins, G W

    2014-07-17

    The recent discovery of more than a thousand planets outside our Solar System, together with the significant push to achieve inertially confined fusion in the laboratory, has prompted a renewed interest in how dense matter behaves at millions to billions of atmospheres of pressure. The theoretical description of such electron-degenerate matter has matured since the early quantum statistical model of Thomas and Fermi, and now suggests that new complexities can emerge at pressures where core electrons (not only valence electrons) influence the structure and bonding of matter. Recent developments in shock-free dynamic (ramp) compression now allow laboratory access to this dense matter regime. Here we describe ramp-compression measurements for diamond, achieving 3.7-fold compression at a peak pressure of 5 terapascals (equivalent to 50 million atmospheres). These equation-of-state data can now be compared to first-principles density functional calculations and theories long used to describe matter present in the interiors of giant planets, in stars, and in inertial-confinement fusion experiments. Our data also provide new constraints on mass-radius relationships for carbon-rich planets.

  12. Voltage spike observation in superconducting cable-in-conduit conductor under ramped magnetic fields. Pt. 1: Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Sangkwon Jeong; Schultz, J.H.; Takayasu, Makoto; Vysotsky, Vitaly; Michael, P.C. [Massachusetts Inst. of Technology, Plasma Fusion Center, Cambridge, MA (United States); Warnes, William [Oregan State Univ., Corvallis, OR (United States); Shen, Stewart [Lawrence Livermore National Lab., Livermore, CA (United States)

    1997-06-01

    A 27-strand hybrid superconducting cable-in-conduit conductor (CICC) was fabricated and tested under quickly-ramped high magnetic fields. When the field increased linearly on the CICC, the voltage signal showed several intermittent spikes before it quenched. This paper describes an observation of peculiar voltage spikes during these ramp-rate limitation experiments. The voltage spikes are interpreted as quench precursors and understood as current redistribution events within the local cable inside the conduit. A quantitative correlation is obtained for the magnetic field at which the first voltage spike occurs during ramping fields. The non-uniform current distribution among the strands and the induced loop current in the cable, which is generated by ramped fields, are found to be responsible for the voltage spikes. (author)

  13. 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.

  14. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    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 burn-up 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. (author)

  15. 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

  16. Burnup analysis of the VVER-1000 reactor using thorium-based fuel

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science

    2014-12-15

    This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.

  17. 40 CFR 1033.520 - Alternative ramped modal cycles.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Alternative ramped modal cycles. 1033.520 Section 1033.520 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR POLLUTION CONTROLS CONTROL OF EMISSIONS FROM LOCOMOTIVES Test Procedures § 1033.520 Alternative ramped...

  18. Inter-Organisational Coordination in Ramp-Up Execution

    DEFF Research Database (Denmark)

    Christensen, Irene; Karlsson, Christer

    While fast product development with early prototyping and reduction of both cycle time and lead time are major concerns, there is little research on ramp up management. This paper examines the structural complexity of the ramp-up processes including the interactions with suppliers and analyses...

  19. Current ramps in tokamaks: from present experiments to ITER scenarios

    NARCIS (Netherlands)

    Imbeaux, F.; Citrin, J.; Hobirk, J.; Hogeweij, G. M. D.; Kochl, F.; Leonov, V. M.; Miyamoto, S.; Nakamura, Y.; Parail, V.; Pereverzev, G.; Polevoi, A.; Voitsekhovitch, I.; Basiuk, V.; Budny, R.; Casper, T.; Fereira, J.; Fukuyama, A.; Garcia, J.; Gribov, Y. V.; Hayashi, N.; Honda, M.; Hutchinson, I. H.; Jackson, G.; Kavin, A. A.; Kessel, C. E.; Khayrutdinov, R. R.; Labate, C.; Litaudon, X.; Lomas, P. J.; Lonnroth, J.; Luce, T.; Lukash, V. E.; Mattei, M.; Mikkelsen, D.; Nunes, I.; Peysson, Y.; Politzer, P.; Schneider, M.; Sips, G.; Tardini, G.; Wolfe, S. M.; Zhogolev, V. E.

    2011-01-01

    In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for

  20. How does the edge height of curb ramps obstruct bicycles?

    Science.gov (United States)

    Hayashi, Masahiro; Uetake, Teruo; Shimoda, Masahiro

    2012-12-01

    The aim of this study is to recommend revisions, based on empirical data, to the current curb ramp standards for keeping bicyclists safe. Four types of curb ramps were tested: (1) concrete with a 50 mm edge height, (2) concrete reinforced by a metal plate with a 50 mm edge height, (3) plastic with a 20 mm edge height, and (4) recycled rubber with a 10 mm edge height. Twenty subjects aged 20-60 years ascended the curbs on a bicycle under various conditions. The angles of approach were 15 degrees, 30 degrees, 45 degrees, 60 degrees, 75 degrees and 90 degrees. Experiments were executed under both wet and dry conditions. We found that when approaching from an angle of 45 degrees or more, all subjects could ascend all ramps under both conditions. From a 15 degrees approach under wet conditions, no subjects ascended the concrete ramps. Some could not ascend at a 15 degrees approach on the concrete ramps in dry conditions, and some could not ascend from a 30 degrees approach on the reinforced concrete ramp in wet conditions. Bicyclists riding on roadways cannot easily ascend a curb ramp with a 50 mm edge, even in dry conditions. We thus recommend that curb ramp edge heights be lower than 50 mm. Keywords: friction coefficient; approach angle

  1. Ramp-rate limitation experiments in support of the TPX magnets

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, S.; Schultz, J.H.; Takayasu, Makoto; Michael, P.C. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Shen, S. [Lawrence Livermore National Lab., CA (United States); Vysotsky, V. [Kurchatov Inst., Moscow (Russian Federation); Warnes, W. [Oregon State Univ., Eugene, OR (United States)

    1995-12-31

    Fast magnetic field change is required for full-size tokamak reactors. The poloidal field magnets are usually ramped to full field at 1.2 T/s, and see pulsed fields of up to 20 T/s during plasma initiation. A new facility has been constructed at M.I.T. that simulates the expected operating conditions of the Tokamak Physics Experiment (TPX) magnets. New features in this facility include (1) a superconducting pulse coil that can superimpose high ramp-down rates, up to 25 T/s, (2 T in 80 msec) on a background field up to 5 T, (2) new power supplies that can supply high rates of dI/dt and dB/dt to the sample under test and the pulse coil, and (3) a forced-flow supercritical helium system that can simulate cooling conditions within the winding pack. The first sample tested in the facility is a 27-strand sub-cable, using 3.1:1 copper, noncopper ratio Nb{sub 3}Sn superconductor, typical of the strands to be used in ten of the poloidal field system magnets. This paper presents the first experimental results on the ramp rate limitation of the sub-size cable sample of TPX PF coil conductor. The transient stability at high ramp rate fields will be discussed.

  2. Sedimentary structures formed by upper-regime flows on a Pleistocene carbonate ramp (Favignana Calcarenite, Sicily, Italy)

    Science.gov (United States)

    Slootman, Arnoud; Moscariello, Andrea; Cartigny, Matthieu; de Boer, Poppe

    2015-04-01

    Antidune, chute-and-pool and cyclic step deposits are found in the outcrops of the Pleistocene calcarenite wedge of Favignana Island. These deposits were formed on a prograding carbonate ramp. Three zones are identified: inner-mid ramp (shoreface), ramp slope, and outer ramp (offshore). The ramp slope dips 3° to 10° and drops 30-40 m over 400-600 m. The ramp slope and outer ramp show a succession of bioturbated dune cross beds with up to 10 m-thick, intercalated event beds containing supercritical-flow structures. Grain sizes range from coarse sand to granules, with large rhodoliths (algal balls) and shells as gravel-sized clasts. It is our aim to provide insight into the processes that create upper-regime flow structures and the hydraulic parameters of their generating flows. During normal storms, wind-driven currents generated submarine dunes that migrated across the sea floor. During exceptional high-energy events (megastorms, tsunamis), large amounts of skeletal debris from the carbonate factory were transported towards the top of the ramp slope, where under the effect of gravity sustained supercritical sediment gravity flows were generated. In a case study of bedform evolution, we present the formation of a large downstream-asymmetric bedform with two antidunes superimposed on its upstream flank. A stepwise flow reconstruction reveals the progressive steepening of the antidunes until critical steepness is reached, and the first and, shortly after, the second antidune wave breaks. The two hydraulic jumps thus formed, developed a temporary cyclic step morphology (i.e. hydraulic jump, accelerating subcritical flow, supercritical chute, hydraulic jump etc.). The bedform geometries are used to reconstruct the nature of the catastrophic events that were active on the ramp slope. The wave length of the antidunes is measured from outcrop, which, through hydraulic equations, allows for estimation of mean flow velocity as a function of sediment concentration in the

  3. Assigning on-ramp flows to maximize highway capacity

    Science.gov (United States)

    Wang, Qiao-Ming; Jiang, Rui; Sun, Xiao-Yan; Wang, Bing-Hong

    2009-09-01

    In this paper, we study the capacity of a highway with two on-ramps by using a cellular automata traffic flow model. We investigate how to improve the system capacity by assigning traffic flow to the two ramps. The system phase diagram is presented and different regions are classified. It is shown that in region I, in which both ramps are in free flow and the main road upstream of the ramps is in congestion, assigning a higher proportion of the demand to the upstream on-ramp could improve the overall flow, which is consistent with previous studies. This is explained through studying the spatiotemporal patterns and analytical investigations. In contrast, optimal assignment has not been observed in other regions. We point out that our result is robust and model independent under certain conditions.

  4. Reattachment heating upstream of short compression ramps in hypersonic flow

    Science.gov (United States)

    Estruch-Samper, David

    2016-05-01

    Hypersonic shock-wave/boundary-layer interactions with separation induce unsteady thermal loads of particularly high intensity in flow reattachment regions. Building on earlier semi-empirical correlations, the maximum heat transfer rates upstream of short compression ramp obstacles of angles 15° ⩽ θ ⩽ 135° are here discretised based on time-dependent experimental measurements to develop insight into their transient nature (Me = 8.2-12.3, Re_h= 0.17× 105-0.47× 105). Interactions with an incoming laminar boundary layer experience transition at separation, with heat transfer oscillating between laminar and turbulent levels exceeding slightly those in fully turbulent interactions. Peak heat transfer rates are strongly influenced by the stagnation of the flow upon reattachment close ahead of obstacles and increase with ramp angle all the way up to θ =135°, whereby rates well over two orders of magnitude above the undisturbed laminar levels are intermittently measured (q'_max>10^2q_{u,L}). Bearing in mind the varying degrees of strength in the competing effect between the inviscid and viscous terms—namely the square of the hypersonic similarity parameter (Mθ )^2 for strong interactions and the viscous interaction parameter bar{χ } (primarily a function of Re and M)—the two physical factors that appear to most globally encompass the effects of peak heating for blunt ramps (θ ⩾ 45°) are deflection angle and stagnation heat transfer, so that this may be fundamentally expressed as q'_max∝ {q_{o,2D}} θ ^2 with further parameters in turn influencing the interaction to a lesser extent. The dominant effect of deflection angle is restricted to short obstacle heights, where the rapid expansion at the top edge of the obstacle influences the relaxation region just downstream of reattachment and leads to an upstream displacement of the separation front. The extreme heating rates result from the strengthening of the reattaching shear layer with the increase in

  5. Key Performance Indicators for the Impact of Cognitive Assembly Planning on Ramp-Up Process

    Directory of Open Access Journals (Sweden)

    Christian Buescher

    2012-01-01

    Full Text Available Within the ramp-up phase of highly automated assembly systems, the planning effort forms a large part of production costs. Due to shortening product lifecycles, changing customer demands, and therefore an increasing number of ramp-up processes, these costs even rise. So assembly systems should reduce these efforts and simultaneously be flexible for quick adaption to changes in products and their variants. A cognitive interaction system in the field of assembly planning systems is developed within the Cluster of Excellence “Integrative production technology for high-wage countries” at RWTH Aachen University which integrates several cognitive capabilities according to human cognition. This approach combines the advantages of automation with the flexibility of humans. In this paper the main principles of the system's core component—the cognitive control unit—are presented to underline its advantages with respect to traditional assembly systems. Based on this, the actual innovation of this paper is the development of key performance indicators. These refer to the ramp-up process as a main objective of such a system is to minimize the planning effort during ramp-up. The KPIs are also designed to show the impact on the main idea of the Cluster of Excellence in resolving the so-called Polylemma of Production.

  6. 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

  7. Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research; Panka, Istvan

    2015-09-15

    Defining precisely the burnup of the nuclear fuel is important from the point of view of core design calculations, safety analyses, criticality calculations (e.g. burnup credit calculations), etc. This paper deals with the uncertainties of MULTICELL calculations obtained by the solution of the OECD NEA UAM PWR pin cell burnup benchmark. In this assessment Monte-Carlo type statistical analyses are applied and the energy dependent covariance matrices of the cross-sections are taken into account. Additionally, the impact of the uncertainties of the fission yields is also considered. The target quantities are the burnup dependent uncertainties of the infinite multiplication factor, the two-group cross-sections, the reaction rates and the number densities of some isotopes up to the burnup of 60 MWd/kgU. In the paper the burnup dependent tendencies of the corresponding uncertainties and their sources are analyzed.

  8. Probabilistic Wind Power Ramp Forecasting Based on a Scenario Generation Method: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Qin [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Florita, Anthony R [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Krishnan, Venkat K [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Hodge, Brian S [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-08-31

    Wind power ramps (WPRs) are particularly important in the management and dispatch of wind power, and they are currently drawing the attention of balancing authorities. With the aim to reduce the impact of WPRs for power system operations, this paper develops a probabilistic ramp forecasting method based on a large number of simulated scenarios. An ensemble machine learning technique is first adopted to forecast the basic wind power forecasting scenario and calculate the historical forecasting errors. A continuous Gaussian mixture model (GMM) is used to fit the probability distribution function (PDF) of forecasting errors. The cumulative distribution function (CDF) is analytically deduced. The inverse transform method based on Monte Carlo sampling and the CDF is used to generate a massive number of forecasting error scenarios. An optimized swinging door algorithm is adopted to extract all the WPRs from the complete set of wind power forecasting scenarios. The probabilistic forecasting results of ramp duration and start time are generated based on all scenarios. Numerical simulations on publicly available wind power data show that within a predefined tolerance level, the developed probabilistic wind power ramp forecasting method is able to predict WPRs with a high level of sharpness and accuracy.

  9. Neutron Transport and Nuclear Burnup Analysis for the Laser Inertial Confinement Fusion-Fission Energy (LIFE) Engine

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E

    2008-10-24

    Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.

  10. Temperature and burnup correlated fuel-cladding chemical interaction in U-10ZR metallic fuel

    Science.gov (United States)

    Carmack, William J.

    Metallic fuels are proposed for use in advanced sodium cooled fast reactors and provide a number of advantages over other fuel types considering their fabricability, performance, recyclability, and safety. Resistance to cladding "breach" and subsequent release of fission products and fuel constituents to the nuclear power plant primary coolant system is a key performance parameter for a nuclear fuel system. In metallic fuel, FCCI weakens the cladding, especially at high power-high temperature operation, contributing to fuel pin breach. Empirical relationships for FCCI have been developed from a large body of data collected from in-pile (EBR-II) and out-of-pile experiments [1]. However, these relationships are unreliable in predicting FCCI outside the range of EBR-II experimental data. This dissertation examines new FCCI data extracted from the MFF-series of prototypic length metallic fuel irradiations performed in the Fast Flux Test Facility (FFTF). The fuel in these assemblies operated a temperature and burnup conditions similar to that in EBR-II but with axial fuel height three times longer than EBR-II experiments. Comparing FCCI formation data from FFTF and EBR-II provides new insight into FCCI formation kinetics. A model is developed combining both production and diffusion of lanthanides to the fuel-cladding interface and subsequent reaction with the cladding. The model allows these phenomena to be influenced by fuel burnup (lanthanide concentrations) and operating temperature. Parameters in the model are adjusted to reproduce measured FCCI layer thicknesses from EBR-II and FFTF. The model predicts that, under appropriate conditions, rate of FCCI formation can be controlled by either fission product transport or by the reaction rate of the interaction species at the fuel-cladding interface. This dissertation will help forward the design of metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full

  11. Quantification of tsunami-induced flows on a Mediterranean carbonate ramp reveals catastrophic evolution

    Science.gov (United States)

    Slootman, Arnoud; Cartigny, Matthieu J. B.; Moscariello, Andrea; Chiaradia, Massimo; de Boer, Poppe L.

    2016-06-01

    Cool-water carbonates are the dominant limestones in the Mediterranean Basin since the Early Pliocene. Their deposition typically resulted in ramp morphologies due to high rates of resedimentation. Several such fossil carbonate ramps are characterised by a bimodal facies stacking pattern, where background deposition of subaqueous dune and/or tempestite deposits is repeatedly interrupted by anomalously thick sedimentary units, dominated by backset-stratification formed by supercritical flows. A multitude of exceptional triggers (e.g. storms, floods, tsunamis) have been invoked to explain the origin of these supercritical flows, which, in the absence of a quantitative analysis, remains speculative as yet. Here, for the first time, the catastrophic evolution of one such Mediterranean carbonate ramp, on Favignana Island (Italy), is quantified by combining 87Sr/86Sr dating, outcrop-based palaeoflow reconstructions and hydraulic calculations. We demonstrate that rare tsunami-induced flows, occurring on average once every 14 to 35 kyr, lasting a few hours only, deposited the anomalously thick backset-bedded units that form half of the sedimentary record. In between such events, cumulative two years of storm-induced flows deposited the remaining half of the succession by the stacking of subaqueous dunes. The two to four orders of magnitude difference in average recurrence period between the two flow types, and their associated sedimentation rates, emphasises the genetic differences between the two styles of deposition. In terms of sediment transport, the studied carbonate ramp was inactive for at least 99% of the time with gradual progradation during decennial to centennial storm activity. Carbonate ramp evolution attained a catastrophic signature by the contribution of rare tsunamis, producing short-lived, high-energy sediment gravity flows.

  12. Study of Freeway Traffic Near an Off-Ramp

    OpenAIRE

    Cassidy, Michael J.; Anani, Shadi B.; Haigwood, John M.

    2000-01-01

    A bottleneck with a diminished capacity is shown to have arisen on a freeway segment whenever queues from the segment's off-ramped spilled over and occupied its mandatory exit lane. It is also shown that longer exit queues from the over-saturate off-ramp were accompanied by lower discharge rates for non-exiting vehicles. The explanation appears to be rubber-necking on the part of the non-exiting drivers. Whenever the of-ramp queues were prevented from spilling over to the exit lane (by changi...

  13. Inter-Organisational Coordination in Ramp-Up Execution

    DEFF Research Database (Denmark)

    Christensen, Irene; Karlsson, Christer

    While fast product development with early prototyping and reduction of both cycle time and lead time are major concerns, there is little research on ramp up management. This paper examines the structural complexity of the ramp-up processes including the interactions with suppliers and analyses...... the degree of fragmentation in the process planning and execution. Resource dependence theory (RDT) is used as central explanatory framework for inter-organisational interdependencies formation throughout the planning and execution of the ramp-up activities and milestones. This study aims at exploring inter...

  14. Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model

    Directory of Open Access Journals (Sweden)

    Abdul Waris

    2008-03-01

    Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.

  15. LHC Report: Ramp-up complete

    CERN Document Server

    Mike Lamont for the LHC Team

    2012-01-01

    The intensity ramp-up outlined in the last Bulletin continued more or less as planned, with some steady running with 1092 bunches followed by the final step-up to 1380 – the maximum number of bunches for the year.   Monday, 16 April was spent performing luminosity calibration. Under carefully controlled conditions and a special beam configuration, the experiments' luminosity measurements were calibrated against a measurement of the accelerator’s absolute luminosity obtained with Van der Meer scans. These scans give the beam sizes at the interaction point which, together with an accurate measurement of the beam and bunch currents (and other more subtle considerations), allow a precise measurement of the absolute luminosity. With the Van der Meer scans out of the way, the final step-up in the number of bunches to 1380 was made. This has resulted in an average peak luminosity of around 5.6x1033 cm-2s-1 in the general-purpose detectors (ATLAS and CMS have yet to publish th...

  16. Ramp length/grade prescriptions for wheelchair dependent individuals.

    Science.gov (United States)

    Canale, I; Felici, F; Marchetti, M; Ricci, B

    1991-09-01

    The aim of this work was to provide well defined criteria for ramp construction for wheelchair dependent individuals (WDI). Force capability was measured in a large sample (140) of WDI, who presented different levels of motor impairment. Levels of impairment were established on the basis of the answers given in a questionnaire regarding the degree of self sufficiency at home as well as outside the home and active participation in sports events. Taking into account those WDI who exhibited at least a minimal level of self-sufficiency, the following prescriptions are indicated. For a 1 metre ramp length, allowable maximal incline 15%; up to 3 metre ramp length, maximal incline 10%. The reliability of such prescriptions was confirmed by having a test ramp traversed by 43 WDI. These values are suggested as confidence limits when faced with public building accommodations. Special prescriptions could be adopted for selected populations of WDI.

  17. Wind Power Ramping Product for Increasing Power System Flexibility

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Mingjian; Zhang, Jie; Wu, Hongyu; Hodge, Bri-Mathias; Ke, Deping; Sun, Yuanzhang

    2016-05-05

    With increasing penetrations of wind power, system operators are concerned about a potential lack of system flexibility and ramping capacity in real-time dispatch stages. In this paper, a modified dispatch formulation is proposed considering the wind power ramping product (WPRP). A swinging door algorithm (SDA) and dynamic programming are combined and used to detect WPRPs in the next scheduling periods. The detected WPRPs are included in the unit commitment (UC) formulation considering ramping capacity limits, active power limits, and flexible ramping requirements. The modified formulation is solved by mixed integer linear programming. Numerical simulations on a modified PJM 5-bus System show the effectiveness of the model considering WPRP, which not only reduces the production cost but also does not affect the generation schedules of thermal units.

  18. Novel Surveillance Technologies for Airport Ramp Area Operations Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the current research is to develop the concept, algorithms and software necessary for enabling a novel surveillance system for airports ramp areas....

  19. Ramp Metering by Limitation of Density and Queue Length

    Science.gov (United States)

    Kamel, Boumediene; Benasser, Amar; Jolly, Daniel

    2009-03-01

    As shown by many works and studies, the control of traffic flow remains the best and the efficient solution of the congestion problems. The ramp metering is a means to reduce congestion effects, it has shown his effectiveness to improve the flow and consequently to reduce the total time spent (TTS) by the vehicles in the network. In Kamel et al. [2008] the concept of flatness-based control was applied to regulate the flow at the mainline section of the road on the traffic flow model LWR. This work was preceded by Abouaïssa et al. [2006]. In this article we have proceeded the same way to set the flat output for a second order model METANET. Unfortunately we were unable to prove that the system is flat and therefore define density as flat output, but we manage to define the control laws in order to respect constraints on density and queue length at the on-ramp. The idea is to express the control variable (the outflow of the on-ramp) according to the output (density or queue length at the on-ramp), then inject it in the system. We use this control method in the ramp metering case. The first aim of the control is to keep the density in the segment where the on-ramp is connected below a density (called the target density YT) for which we have the lowest TTS, and the second aim is to keep the queue length at the on-ramp below a maximum value. In order to evaluate the controller's efficiency and applicability, a comparison is made with traditional ALINEA based controller. We illustrate this approach by comparing the cases 'no control' and 'ramp metering' for a simple network.

  20. Analytical delay models for RLC interconnects under ramp input

    Institute of Scientific and Technical Information of China (English)

    REN Yinglei; MAO Junfa; LI Xiaochun

    2007-01-01

    Analytical delay models for Resistance Inductance Capacitance (RLC)interconnects with ramp input are presented for difierent situations,which include overdamped,underdamped and critical response cases.The errors of delay estimation using the analytical models proposed in this paper are less bv 3%in comparison to the SPICE-computed delay.These models are meaningful for the delay analysis of actual circuits in which the input signal is ramp but not ideal step input.

  1. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  2. First steps towards a validation of the new burnup and depletion code TNT

    Energy Technology Data Exchange (ETDEWEB)

    Herber, S.C.; Allelein, H.J. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6); Friege, N. [RWTH Aachen (Germany). Inst. for Reactor Safety and Reactor Technology; Kasselmann, S. [Research Center Juelich (Germany). Inst. for Energy and Climate Research - Nuclear Waste Disposal and Reactor Safety (IEK-6)

    2012-11-01

    In the frame of the fusion of the core design calculation capabilities, represented by V.S.O.P., and the accident calculation capabilities, represented by MGT(-3D), the successor of the TINTE code, difficulties were observed in defining an interface between a program backbone and the ORIGEN code respectively the ORIGENJUEL code. The estimation of the effort of refactoring the ORIGEN code or to write a new burnup code from scratch, led to the decision that it would be more efficient writing a new code, which could benefit from existing programming and software engineering tools from the computer code side and which can use the latest knowledge of nuclear reactions, e.g. consider all documented reaction channels. Therefore a new code with an object-oriented approach was developed at IEK-6. Object-oriented programming is currently state of the art and provides mostly an improved extensibility and maintainability. The new code was named TNT which stands for Topological Nuclide Transformation, since the code makes use of the real topology of the nuclear reactions. Here we want to present some first validation results from code to code benchmarks with the codes ORIGEN V2.2 and FISPACT2005 and whenever possible analytical results also used for the comparison. The 2 reference codes were chosen due to their high reputation in the field of fission reactor analysis (ORIGEN) and fusion facilities (FISPACT). (orig.)

  3. PDR with a foot-mounted IMU and ramp detection.

    Science.gov (United States)

    Jiménez, Antonio R; Seco, Fernando; Zampella, Francisco; Prieto, José C; Guevara, Jorge

    2011-01-01

    The localization of persons in indoor environments is nowadays an open problem. There are partial solutions based on the deployment of a network of sensors (Local Positioning Systems or LPS). Other solutions only require the installation of an inertial sensor on the person's body (Pedestrian Dead-Reckoning or PDR). PDR solutions integrate the signals coming from an Inertial Measurement Unit (IMU), which usually contains 3 accelerometers and 3 gyroscopes. The main problem of PDR is the accumulation of positioning errors due to the drift caused by the noise in the sensors. This paper presents a PDR solution that incorporates a drift correction method based on detecting the access ramps usually found in buildings. The ramp correction method is implemented over a PDR framework that uses an Inertial Navigation algorithm (INS) and an IMU attached to the person's foot. Unlike other approaches that use external sensors to correct the drift error, we only use one IMU on the foot. To detect a ramp, the slope of the terrain on which the user is walking, and the change in height sensed when moving forward, are estimated from the IMU. After detection, the ramp is checked for association with one of the existing in a database. For each associated ramp, a position correction is fed into the Kalman Filter in order to refine the INS-PDR solution. Drift-free localization is achieved with positioning errors below 2 meters for 1,000-meter-long routes in a building with a few ramps.

  4. Dynamic control for nanostructures through slowly ramping parameters

    Science.gov (United States)

    Yoo, Jaeyun; Blick, Robert; Ahn, Kang-Hun

    2016-06-01

    We propose a nanostructure control method which uses slowly ramping parameters. We demonstrate the dynamics of this method in both a nonlinear classical system and a quantum system. When a quantum mechanical two-level atom (quantum dot) is irradiated by an electric field with a slowly increasing frequency, there exists a sudden transition from ground (excited) to excited (ground) state. This occurs when the ramping rate is smaller than the square of the Rabi frequency. The transition arises when its "instant frequency"—the time derivative of the driving field phase—matches the resonance frequency, satisfying the Fermi golden rule. We also find that the parameter ramping is an efficient control manner for classical nanomechanical shuttles. For ramping of driving amplitudes, the shuttle's mechanical oscillation is amplified and even survives when the ramping is stopped outside the original oscillation region. This strange oscillation is due to the entrance into a multistable dynamic region in phase space. For ramping of driving frequencies, an onset of oscillation arises when the instant frequency enters the oscillation region. Thus, regardless of being classical or quantum, the instant frequency is physically relevant. We discuss in which conditions the dynamic control is efficient.

  5. 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

  6. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  7. 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.

  8. 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.

  9. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  10. 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.

  11. Numerical Study of Control of Flow Separation Over a Ramp with Nanosecond Plasma Actuator

    Science.gov (United States)

    Zheng, J. G.; Khoo, B. C.; Cui, Y. D.; Zhao, Z. J.; Li, J.

    2016-06-01

    The nanosecond plasma discharge actuator driven by high voltage pulse with typical rise and decay time of several to tens of nanoseconds is emerging as a promising active flow control means in recent years and is being studied intensively. The characterization study reveals that the discharge induced shock wave propagates through ambient air and introduces highly transient perturbation to the flow. On the other hand, the residual heat remaining in the discharge volume may trigger the instability of external flow. In this study, this type of actuator is used to suppress flow separation over a ramp model. Numerical simulation is carried out to investigate the interaction of the discharge induced disturbance with the external flow. It is found that the flow separation region over the ramp can be reduced significantly. Our work may provide some insights into the understanding of the control mechanism of nanosecond pulse actuator.

  12. Extracting Strength from Ramp-Release Experiments on Z

    Science.gov (United States)

    Brown, Justin

    2013-06-01

    Releasing from a compressed state has long been recognized as a sensitive measure of a material's constitutive response. The initial elastic unloading provides insights which can be related to changes in shear stress or, in the context of classic plasticity, to the material's yield surface. Ramp compression and subsequent release experiments on Sandia's Z machine typically consist of a driving aluminum electrode pushing a sample material which is backed by a window. A particle velocity measurement of the sample/window interface provides a ramp-release profile. Under most circumstances, however, the impedance mismatch at this interface results in the measurement of a highly perturbed velocity, particularly at the late times of interest. Wave attenuation, the finite pressure range over which the material elastically unloads, and rate effects additionally complicate the interpretation of the experiment. In an effort to accurately analyze experiments of this type, each of these complications is addressed. The wave interactions are accounted for through the so-called transfer function methodology and involves a coupling of the experimental measurements with numerical simulations. Simulated window velocity measurements are combined with the corresponding in situ simulations to define a mapping describing the wave interactions due to the presence of the window. Applying this mapping to the experimentally measured velocity results in an in situ sample response which may then be used in a classic Lagrangian analysis from which the strength can be extracted via the self-consistent method. Corrections for attenuation, pressure averaging, and limitations of the analysis due to rate-effects are verified through the use of synthetic data. To date, results on the strength of aluminum to 1.2 MBar, beryllium to 1 MBar, and tantalum to over 2 MBar have been obtained through this methodology and will be presented. Sandia National Laboratories is a multi-program laboratory operated by

  13. Verification of spectral burn-up codes on 2D fuel assemblies of the GFR demonstrator ALLEGRO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Čerba, Štefan, E-mail: stefan.cerba@stuba.sk [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Vrban, Branislav; Lüley, Jakub [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia); Dařílek, Petr [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Zajac, Radoslav, E-mail: radoslav.zajac@vuje.sk [VUJE a.s., Okružná 5, 918 64 Trnava (Slovakia); Nečas, Vladimír; Haščik, Ján [Slovak University of Technology in Bratislava, Faculty of Electrical Engineering and Information Technology, Institute of Nuclear and Physical Engineering, Ilkovičova 3, 812 19 Bratislava (Slovakia)

    2014-02-15

    Highlights: • Verification of the MCNPX, HELIOS and SCALE codes. • MOX and ceramic fuel assembly. • Gas-cooled fast reactor. • Burnup calculation. - Abstract: The gas-cooled fast reactor, which is one of the six GEN IV reactor concepts, is characterized by high operational temperatures and a hard neutron spectrum. The utilization of commonly used spectral codes, developed mainly for LWR reactors operated in the thermal/epithermal neutron spectrum, may be connected with systematic deviations since the main development effort of these codes has been focused on the thermal part of the neutron spectrum. To be able to carry out proper calculations for fast systems the used codes have to account for neutron resonances including the self-shielding effect. The presented study aims at verifying the spectral HELIOS, MCNPX and SCALE codes on the basis of depletion calculations of 2D MOX and ceramic fuel assemblies of the ALLEGRO gas-cooled fast reactor demonstrator in infinite lattice.

  14. Power Ramp Limitation capabilities of Large PV Power Plants with Active Power Reserves

    DEFF Research Database (Denmark)

    Bogdan, Craciun; Kerekes, Tamas; Sera, Dezso

    2017-01-01

    Power Ramp Limitation (PRL) is likely to become a requirement for large scale photovoltaic power plants (LPVPPs) in order to allow the increase of PV penetration levels. Especially in islands with reduced inertia capability, this problem is more stringent: high power ramp can be caused by either...... fast irradiance changes or other participant generators for example wind power, or loads. In order to compensate for the power mismatch, LPVPPs must use Active Power Reserve (APR), by either curtailment or auxiliary storage. The paper proposes a PRL control structure for dynamic APR sizing...... area of LPVPPs acts as filter against fast irradiance changes, the study reveals also the required plant size for which auxiliary storage is no longer needed in order to comply with PRL requirements – an important economical aspect....

  15. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  16. 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.

  17. Reactivity effect of spent fuel depending on burn-up history

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Takafumi [Nagoya Univ., Nagoya, Aichi (Japan); Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Mochizuki, Hiroki [The Japan Research Institute, Ltd., Tokyo (Japan)

    2001-06-01

    It is well known that a composition of spent fuel depends on various parameter changes throughout a burn-up period. In this study we aimed at the boron concentration and its change, the coolant temperature and its spatial distribution, the specific power, the operation mode, and the duration of inspection, because the effects due to these parameters have not been analyzed in detail. The composition changes of spent fuel were calculated by using the burn-up code SWAT, when the parameters mentioned above varied in the range of actual variations. Moreover, to estimate the reactivity effect caused by the composition changes, the criticality calculations for an infinite array of spent fuel were carried out with computer codes SRAC95 or MVP. In this report the reactivity effects were arranged from the viewpoint of what parameters gave more positive reactivity effect. The results obtained through this study are useful to choose the burn-up calculation model when we take account of the burn-up credit in the spent fuel management. (author)

  18. Comparison between SERPENT and MONTEBURNS codes applied to burnup calculations of a GFR-like configuration

    Energy Technology Data Exchange (ETDEWEB)

    Chersola, Davide [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Marotta, Riccardo [GeNERG – DIME/TEC, University of Genova, via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, via Dodecaneso 33, 16146 Genova (Italy); Mazzini, Guido [Centrum výzkumu Řež (Research Centre Rez), Husinec-Rez, cp. 130, 25068 Rez (Czech Republic)

    2014-07-01

    Highlights: • MC codes are widely adopted to analyze nuclear facilities, including GEN-IV reactors. • Burnup calculations are an efficient tool to test neutronic Monte Carlo codes. • In this comparison the used codes show some differences but a good agreement exists. - Abstract: This paper presents the comparison between two Monte Carlo based burnup codes: SERPENT and MONTEBURNS. Monte Carlo codes are fully and worldwide adopted to perform analyses on nuclear facilities, also in the frame of Generation IV advanced reactors simulations. Thus, faster and most powerful calculation codes are needed with the aim to analyze complex geometries and specific neutronic behaviors. Burnup calculations are an efficient tool to test neutronic Monte Carlo codes: indeed these calculations couple transport and depletion procedures, so that neutronic reactor behavior can be simulated in its totality. Comparisons have been performed on a configuration representing the Allegro MOX 75 MW{sub th} reactor proposed by the European GoFastR (Gas-cooled Fast Reactor) Project in the frame of the 7th Euratom Framework Program. Although in burnup and criticality comparisons the codes used in simulations show different calculation times and some differences in amounts and in precision (in term of statistical errors), a reasonably good agreement between them exists.

  19. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [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: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  20. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-15

    Highlights: • Thermal properties of irradiated U–Mo alloy monolithic fuel samples were measured. • Density, thermal diffusivity, and thermal conductivity are influenced by increasing burnup. • U–Mo chemistry and specific heat capacity was not as sensitive to increasing burnup. • Thermal conductivity decreased approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} at 200 °C. • An empirical model developed previously agrees well with the experimental measurements. - Abstract: A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U–Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U–Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U–Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U–Mo alloy decreased approximately 30% for a fission density of 3.30 × 10{sup 21} fissions cm{sup −3} and approximately 45% for a fission density of 4.52 × 10{sup 21} fissions cm{sup −3} from unirradiated values at 200 °C. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  1. 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.

  2. Meniscal Ramp Lesions: Anatomy, Incidence, Diagnosis, and Treatment.

    Science.gov (United States)

    Chahla, Jorge; Dean, Chase S; Moatshe, Gilbert; Mitchell, Justin J; Cram, Tyler R; Yacuzzi, Carlos; LaPrade, Robert F

    2016-07-01

    Meniscal ramp lesions are more frequently associated with anterior cruciate ligament (ACL) injuries than previously recognized. Some authors suggest that this entity results from disruption of the meniscotibial ligaments of the posterior horn of the medial meniscus, whereas others support the idea that it is created by a tear of the peripheral attachment of the posterior horn of the medial meniscus. Magnetic resonance imaging (MRI) scans have been reported to have a low sensitivity, and consequently, ramp lesions often go undiagnosed. Therefore, to rule out a ramp lesion, an arthroscopic evaluation with probing of the posterior horn of the medial meniscus should be performed. Several treatment options have been reported, including nonsurgical management, inside-out meniscal repair, or all-inside meniscal repair. In cases of isolated ramp lesions, a standard meniscal repair rehabilitation protocol should be followed. However, when a concomitant ACL reconstruction (ACLR) is performed, the rehabilitation should follow the designated ACLR postoperative protocol. The purpose of this article was to review the current literature regarding meniscal ramp lesions and summarize the pertinent anatomy, biomechanics, diagnostic strategies, recommended treatment options, and postoperative protocol.

  3. A RAMP CODE FOR FINE-GRAINED ACCESS CONTROL

    Directory of Open Access Journals (Sweden)

    Kannan Karthik

    2013-02-01

    Full Text Available Threshold ramp secret sharing schemes are designed so that (i certain subsets of shares have no information about the secret, (ii some subsets have partial information about the secret and (iii some subsets have complete information to recover the secret. However most of the ramp schemes in present literature do not control the leakage of information in partial access sets, due to which the information acquired by these sets is devoid of structure and not useful for fine-grained access control. Through a non-perfect secret sharing scheme called MIX-SPLIT, an encoding methodology for controlling the leakage in partial access sets is proposed and this is used for fine-grained access to binary strings. The ramp code generated using MIX-SPLIT requires a much smaller share size of O(n, as compared to Shamir's ramp adaptation which incurs a share size of atleast O(n2 for the same multi-access structure. The proposed ramp code is finally applied towards the protection and fine-grained access of industrial design drawings.

  4. Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.

    2001-09-28

    The Interim Staff Guidance on burnup credit (ISG-8) issued by the United States Nuclear Regulatory Commission's (U.S. NRC) Spent Fuel Project Office recommends restricting the use of burnup credit to assemblies that have not used burnable absorbers. This recommended restriction eliminates a large portion of the currently discharged spent fuel assemblies from cask loading, and thus severely limits the practical usefulness of burnup credit. In the absence of readily available information on burnable poison rod (BPR) design specifications and usage in U.S. pressurized-water-reactors (PWRs), and the subsequent reactivity effect of BPR exposure on discharged spent nuclear fuel (SNF), NRC staff has indicated a need for additional information in these areas. In response, this report presents a parametric study of the effect of BPR exposure on the reactivity of SNF for various BPR designs, fuel enrichments, and exposure conditions, and documents BPR design specifications. Trends in the reactivity effects of BPRs are established with infinite pin-cell and assembly array calculations with the SCALE and HELIOS code packages, respectively. Subsequently, the reactivity effects of BPRs for typical initial enrichment and burnup combinations are quantified based on three-dimensional (3-D) KENO V.a Monte Carlo calculations with a realistic rail-type cask designed for burnup credit. The calculations demonstrate that the positive reactivity effect due to BPR exposure increases nearly linearly with burnup and is dependent on the number, poison loading, and design of the BPRs and the initial fuel enrichment. Expected typical reactivity increases, based on one-cycle BPR exposure, were found to be less than 1% {Delta}k. Based on the presented analysis, guidance is offered on an appropriate approach for calculating bounding SNF isotopic data for assemblies exposed to BPRs. Although the analyses do not address the issue of validation of depletion methods for assembly designs with BPRs

  5. Interface resistance of YBa2Cu3O7−δ/La0.67Sr0.33MnO3 ramp-type contacts

    NARCIS (Netherlands)

    Zalk, van M.; Brinkman, A.; Aarts, J.; Hilgenkamp, H.

    2010-01-01

    We fabricated and characterized YBa2Cu3O7−δ/La0.67Sr0.33MnO3 (YBCO/LSMO) ramp-type contacts and junctions. An interlayer technique was applied to repair the ramp stoichiometry after etching. It was found that, typically, the resistance of the YBCO/LSMO interface is high compared to the resistances o

  6. A Numerical Model to Simulate the Motion of a Lifesaving Module During its Launching from the Ship’S Stern Ramp

    Directory of Open Access Journals (Sweden)

    Dymarski Paweł

    2014-04-01

    Full Text Available The article presents a numerical model of object motion in six degrees of freedom (DoF which is intended to be used to simulate 3D motion of a lifesaving module during its launching from a ship using a stern ramp in rough sea. The model, of relatively high complexity, takes into account both the motion of the ship on water in changing sea conditions, and the relative motion of the ramp with respect to the ship.

  7. Resident-Assisted Montessori Programming (RAMP): training persons with dementia to serve as group activity leaders.

    Science.gov (United States)

    Camp, Cameron J; Skrajner, Michael J

    2004-06-01

    The purpose of this study was to determine the effects of an activity implemented by means of Resident-Assisted Montessori Programming (RAMP). Four persons with early-stage dementia were trained to serve as leaders for a small-group activity played by nine persons with more advanced dementia. Assessments of leaders' ability to learn the procedures of leading a group, as well as their satisfaction with this role, were taken, as were measures of players' engagement and affect during standard activities programming and RAMP activities. Leaders demonstrated the potential to fill the role of group activity leader effectively, and they expressed a high level of satisfaction with this role. Players' levels of positive engagement and pleasure during the RAMP activity were higher than during standard group activities. This study suggests that to the extent that procedural learning is available to persons with early-stage dementia, especially when they are assisted with external cueing, these individuals can successfully fill the role of volunteers when working with persons with more advanced dementia. This can provide a meaningful social role for leaders and increase access to high quality activities programming for large numbers of persons with dementia. Copyright 2004 The Gerontological Society of America

  8. Enhancing Power System Operational Flexibility With Flexible Ramping Products: A Review

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Qin; Hodge, Bri-Mathias

    2017-08-01

    With the increased variability and uncertainty of net load induced from high penetrations of renewable energy resources and more flexible interchange schedules, power systems are facing great operational challenges in maintaining balance. Among these, the scarcity of ramp capability is an important culprit of power balance violations and high scarcity prices. To address this issue, market-based flexible ramping products (FRPs) have been proposed in the industry to improve the availability of ramp capacity. This paper presents an in-depth review of the modeling and implementation of FRPs. The major motivation is that although FRPs are widely discussed in the literature, it is still unclear to many how they can be incorporated into a co-optimization framework that includes energy and ancillary services. The concept and a definition of power system operational flexibility as well as the needs for FRPs are introduced. The industrial practices of implementing FRPs under different market structures are presented. Market operation issues and future research topics are also discussed. This paper can provide researchers and power engineers with further insights into the state of the art, technical barriers, and potential directions for FRPs.

  9. The Dynamic Behaviors of Single Crystal RDX Under Ramp Wave Loading to 15GPa

    Science.gov (United States)

    Wang, Guiji; Cai, Jintao; Zhao, Jianheng; Zhao, Feng; Wu, Gang; Tan, Fuli; Sun, Chengwei

    Based on high pulsed power generator CQ-4, the single crystal RDX explosive was researched along different crystal orientations under ramp wave loadings up to 15 GPa. The typical three-wave structures were obtained by means of laser interferometry PDV, which show the elastic-plastic transition and α to γ phase transition. The ramp elastic limit (REL) and yield strength of RDX along 210 and 100 crystal orientations were respectively calculated and the resuts show obvious effects of crystal orientaions for RDX. The ramp elastic limit σIEL of RDX along 210 orientation is 0.688-0.758GPa, and the σIEL of RDX along 100 is 1.039 -1.110 GPa. The α to γ phase transformation characteristics were also analyzed based on the experimental data. The initial phase transition pressure for the two crystal orientation of RDX are about 3.5 to 4 GPa, which agree well with the data of about 4-5GPa given by MD simulation. The data directly validate the results given by Raman Spectrum under shock compression and static high pressure, which couldn't be observed by wave profiles. The experimental data can be used to verify and validate the new models of RDX under dynamic loading. Supported by NSFC of China under Contract No.11327803 and 11176002

  10. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group 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 are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  11. 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

  12. Speed limit and ramp meter control for traffic flow networks

    Science.gov (United States)

    Goatin, Paola; Göttlich, Simone; Kolb, Oliver

    2016-07-01

    The control of traffic flow can be related to different applications. In this work, a method to manage variable speed limits combined with coordinated ramp metering within the framework of the Lighthill-Whitham-Richards (LWR) network model is introduced. Following a 'first-discretize-then-optimize' approach, the first order optimality system is derived and the switch of speeds at certain fixed points in time is explained, together with the boundary control for the ramp metering. Sequential quadratic programming methods are used to solve the control problem numerically. For application purposes, experimental setups are presented wherein variable speed limits are used as a traffic guidance system to avoid traffic jams on highway interchanges and on-ramps.

  13. PDR with a Foot-Mounted IMU and Ramp Detection

    Directory of Open Access Journals (Sweden)

    Jorge Guevara

    2011-09-01

    Full Text Available The localization of persons in indoor environments is nowadays an open problem. There are partial solutions based on the deployment of a network of sensors (Local Positioning Systems or LPS. Other solutions only require the installation of an inertial sensor on the person’s body (Pedestrian Dead-Reckoning or PDR. PDR solutions integrate the signals coming from an Inertial Measurement Unit (IMU, which usually contains 3 accelerometers and 3 gyroscopes. The main problem of PDR is the accumulation of positioning errors due to the drift caused by the noise in the sensors. This paper presents a PDR solution that incorporates a drift correction method based on detecting the access ramps usually found in buildings. The ramp correction method is implemented over a PDR framework that uses an Inertial Navigation algorithm (INS and an IMU attached to the person’s foot. Unlike other approaches that use external sensors to correct the drift error, we only use one IMU on the foot. To detect a ramp, the slope of the terrain on which the user is walking, and the change in height sensed when moving forward, are estimated from the IMU. After detection, the ramp is checked for association with one of the existing in a database. For each associated ramp, a position correction is fed into the Kalman Filter in order to refine the INS-PDR solution. Drift-free localization is achieved with positioning errors below 2 meters for 1,000-meter-long routes in a building with a few ramps.

  14. 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

  15. Burnup calculation by the method of first-flight collision probabilities using average chords prior to the first collision

    Science.gov (United States)

    Karpushkin, T. Yu.

    2012-12-01

    A technique to calculate the burnup of materials of cells and fuel assemblies using the matrices of first-flight neutron collision probabilities rebuilt at a given burnup step is presented. A method to rebuild and correct first collision probability matrices using average chords prior to the first neutron collision, which are calculated with the help of geometric modules of constructed stochastic neutron trajectories, is described. Results of calculation of the infinite multiplication factor for elementary cells with a modified material composition compared to the reference one as well as calculation of material burnup in the cells and fuel assemblies of a VVER-1000 are presented.

  16. 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

  17. Coupling Measurements and Corrections for the Combined Ramp and Squeeze

    CERN Document Server

    Persson, Tobias Hakan Bjorn; Langner, Andy Sven; Malina, Lukas; Maclean, Ewen Hamish; Coello De Portugal - Martinez Vazquez, Jaime Maria; Redaelli, Stefano; Salvachua Ferrando, Belen Maria; Schaumann, Michaela; Solfaroli Camillocci, Matteo; Skowronski, Piotr Krzysztof; Tomas Garcia, Rogelio; Garcia-Tabares Valdivieso, Ana; Wenninger, Jorg; CERN. Geneva. ATS Department

    2016-01-01

    In operation of the LHC the ramp and the squeeze process have been independent beam processes up to now. Making them into a combined process would save time to reach the point where the beams are brought to collision. This would increase the integrated luminosity provided by the LHC. One possible source of problems could be deviation from the ideal optics and in particular the control of the transverse coupling. In this report we focus on the coupling measurements that were taken during the Combined Ramp and Squeeze (CRS) MD.

  18. 2-D Electromagnetic Model of Fast-Ramping Superconducting Magnets

    CERN Document Server

    Auchmann, B; Kurz, S; Russenschuck, Stephan

    2006-01-01

    Fast-ramping superconducting (SC) accelerator magnets are the subject of R&D efforts by magnet designers at various laboratories. They require modifications of magnet design tools such as the ROXIE program at CERN, i.e. models of dynamic effects in superconductors need to be implemented and validated. In this paper we present the efforts towards a dynamic 2-D simulation of fast-ramping SC magnets with the ROXIE tool. Models are introduced and simulation results are compared to measurements of the GSI001 magnet of a GSI test magnet constructed and measured at BNL.

  19. Observation of the AC Josephson effect up to THZ frequencies in YBCO/PBCO/YBCO ramp-type Josephson junctions

    NARCIS (Netherlands)

    Terpstra, D.; Rijnders, A.J.H.M.; Roesthuis, F.J.G.; Blank, D.H.A.; Gerritsma, G.J.; Rogalla, H.

    1993-01-01

    We present the response to 100 GHz irradiation of high-Tc Josephson junction devices for mixer/detector applications in the (sub-) mm wave range. These devices consist of a YBCO/PBCO/YBCO ramp-type junction combined with a planar logarithmic periodic antenna. The critical current and the first two S

  20. Development of a Burnup Module DECBURN Based on the Krylov Subspace Method

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S

    2008-05-15

    This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution

  1. Experiencing Production Ramp-Up Education for Engineers

    Science.gov (United States)

    Bassetto, S.; Fiegenwald, V.; Cholez, C.; Mangione, F.

    2011-01-01

    This paper presents a game of industrialisation, based on a paper airplane, that mimics real world production ramp-up and blends classical engineering courses together. It is based on a low cost product so that it can be mass produced. The game targets graduate students and practitioners in engineering fields. For students, it offers an experiment…

  2. Cored Rutherford cables for the GSI fast ramping synchrotron

    NARCIS (Netherlands)

    Wilson, M.N.; Ghosh, A.K.; Haken, ten B.; Hassenzahl, W.V.; Kaugerts, J.; Moritz, G.; Muehle, C.; Ouden, den A.; Soika, R.; Wanderer, P.; Wessel, W.A.J.

    2003-01-01

    The new heavy ion synchrotron facility proposed by GSI will have two superconducting magnet rings in the same tunnel, with rigidities of 200 T/spl middot/m and 100 T/spl middot/m. Fast ramp times are needed, which can cause significant problems for the magnets, particularly in the areas of ac loss a

  3. Building Ramps and Hovercrafts and Improving Math Skills.

    Science.gov (United States)

    Bottge, Brian A.

    2001-01-01

    This article describes a video- and computer-based program used to motivate and develop mathematics skills in middle school students with disabilities. The program emphasizes real-life problems such as building a cage for a pet, a skate boarding ramp, and a "hovercraft" frame. Case studies illustrate the program's effectiveness with individual…

  4. 9 CFR 313.1 - Livestock pens, driveways and ramps.

    Science.gov (United States)

    2010-01-01

    ... animal may be injured shall be repaired. (b) Floors of livestock pens, ramps, and driveways shall be constructed and maintained so as to provide good footing for livestock. Slip resistant or waffled floor... the opinion of the inspector, to protect them from the adverse climatic conditions of the locale...

  5. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aufiero, M.; Cammi, A.; Fiorina, C. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Leppänen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Luzzi, L., E-mail: lelio.luzzi@polimi.it [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy); Ricotti, M.E. [Politecnico di Milano, Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via Ponzio, 34/3, I-20133 Milano (Italy)

    2013-10-15

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  6. An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor

    Science.gov (United States)

    Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.

    2013-10-01

    In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR). This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation. The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution. Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions. The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.

  7. Numerical investigation on jet interaction with a compression ramp

    Institute of Scientific and Technical Information of China (English)

    Zhen Huaping; Gao Zhenxun; Lee Chunhian

    2013-01-01

    A numerical investigation on jet interaction in supersonic laminar flow with a compression ramp is performed utilizing the AUSMDV scheme and a parallel solver.Several parameters dominating the interference flowfield are studied after defining the relative increment of normal force and the jet amplification factor as the evaluation criterion of jet control performance.The computational results show that most features of the interaction flowfield between the transverse jet and the ramp are similar to those between a jet and a flat plate,except that the flow structures are more complicated and the low-pressure region behind the jet is less extensive.The relative force increment and the jet amplification factor both increase with the distance between the jet and the ramp shortening till quintuple jet diameters.Inconspicuous difference is observed between the jet-before-ramp and jet-on-ramp cases.The variation of the injection angle changes the extent of the separation region,the plateau pressure,and the peak pressure near the jet.In the present computational conditions,120° is indicated relatively optimal among all the injection angles studied.For cold gas simulations,although little influence of the jet temperature on the pressure distribution near the jet is observed under the computation model and the flow parameters studied,reducing jet temperature somehow benefits the improvement of the normal force and the jet efficiency.When the pressure ratio of jet to freestream is fixed,the relative force increment varies little when increasing the freestream Mach number,while the jet amplification factor increases.

  8. Determination of deuterium–tritium critical burn-up parameter by four temperature theory

    Energy Technology Data Exchange (ETDEWEB)

    Nazirzadeh, M.; Ghasemizad, A. [Department of Physics, University of Guilan, 41335-1914 Rasht (Iran, Islamic Republic of); Khanbabei, B. [School of Physics, Damghan University, 36716-41167 Damghan (Iran, Islamic Republic of)

    2015-12-15

    Conditions for thermonuclear burn-up of an equimolar mixture of deuterium-tritium in non-equilibrium plasma have been investigated by four temperature theory. The photon distribution shape significantly affects the nature of thermonuclear burn. In three temperature model, the photon distribution is Planckian but in four temperature theory the photon distribution has a pure Planck form below a certain cut-off energy and then for photon energy above this cut-off energy makes a transition to Bose-Einstein distribution with a finite chemical potential. The objective was to develop four temperature theory in a plasma to calculate the critical burn up parameter which depends upon initial density, the plasma components initial temperatures, and hot spot size. All the obtained results from four temperature theory model are compared with 3 temperature model. It is shown that the values of critical burn-up parameter calculated by four temperature theory are smaller than those of three temperature model.

  9. 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.

  10. Thermal properties of U–Mo alloys irradiated to moderate burnup and power

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Casella, Andrew M.; Casella, Amanda J.; Buck, Edgar C.; Pool, Karl N.; MacFarlan, Paul J.; Edwards, Matthew K.; Smith, Frances N.

    2015-09-01

    A variety of physical and thermal property measurements as a function of temperature and fission density were performed on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in aluminum alloy 6061. The U-Mo alloy density, thermal diffusivity, and thermal conductivity are strongly influenced by increasing burnup, mainly as the result of irradiation induced recrystallization and fission gas bubble formation and coalescence. U-Mo chemistry, specifically Mo content, and specific heat capacity was not as sensitive to increasing burnup. Measurements indicated that thermal conductivity of the U-Mo alloy decreased approximately 30% for a fission density of 2.88 × 1021 fissions cm-3 and approximately 45% for a fission density of 4.08 × 1021 fissions cm-3 from unirradiated values at 200 oC. An empirical thermal conductivity degradation model developed previously and summarized here agrees well with the experimental measurements.

  11. Assessing the Effect of Fuel Burnup on Control Rod Worth for HEU and LEU Cores of Gharr-1

    Directory of Open Access Journals (Sweden)

    E.K. Boafo

    2013-02-01

    Full Text Available An important parameter in the design and analysis of a nuclear reactor is the reactivity worth of the control rod which is a measure of the efficiency of the control rod to absorb excess reactivity. During reactor operation, the control rod worth is affected by factors such as the fuel burnup, Xenon concentration, Samarium concentration and the position of the control rod in the core. This study investigates the effect of fuel burnup on the control rod worth by comparing results of a fresh and an irradiated core of Ghana's Miniature Neutron Source Reactor for both HEU and LEU cores. In this study, two codes have been utilized namely BURNPRO for fuel burnup calculation and MCNP5 which uses densities of actinides of the irradiated fuel obtained from BURNPRO. Results showed a decrease of the control rod worth with burnup for the LEU while rod worth increased with burnup for the HEU core. The average thermal flux in both inner and outer irradiation sites also decreased significantly with burnup for both cores.

  12. S∧4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

    Science.gov (United States)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The S∧4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at ~1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S∧4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 ¢/K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S∧4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 ¢/K -0.9073 and 0.8502 $/atom%) are the lowest.

  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. 30 CFR 56.9303 - Construction of ramps and dumping facilities.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Construction of ramps and dumping facilities... Loading, Hauling, and Dumping Safety Devices, Provisions, and Procedures for Roadways, Railroads, and Loading and Dumping Sites § 56.9303 Construction of ramps and dumping facilities. Ramps and...

  15. Improving short-term forecasting during ramp events by means of Regime-Switching Artificial Neural Networks

    Science.gov (United States)

    Gallego, C.; Costa, A.; Cuerva, A.

    2010-09-01

    Since nowadays wind energy can't be neither scheduled nor large-scale storaged, wind power forecasting has been useful to minimize the impact of wind fluctuations. In particular, short-term forecasting (characterised by prediction horizons from minutes to a few days) is currently required by energy producers (in a daily electricity market context) and the TSO's (in order to keep the stability/balance of an electrical system). Within the short-term background, time-series based models (i.e., statistical models) have shown a better performance than NWP models for horizons up to few hours. These models try to learn and replicate the dynamic shown by the time series of a certain variable. When considering the power output of wind farms, ramp events are usually observed, being characterized by a large positive gradient in the time series (ramp-up) or negative (ramp-down) during relatively short time periods (few hours). Ramp events may be motivated by many different causes, involving generally several spatial scales, since the large scale (fronts, low pressure systems) up to the local scale (wind turbine shut-down due to high wind speed, yaw misalignment due to fast changes of wind direction). Hence, the output power may show unexpected dynamics during ramp events depending on the underlying processes; consequently, traditional statistical models considering only one dynamic for the hole power time series may be inappropriate. This work proposes a Regime Switching (RS) model based on Artificial Neural Nets (ANN). The RS-ANN model gathers as many ANN's as different dynamics considered (called regimes); a certain ANN is selected so as to predict the output power, depending on the current regime. The current regime is on-line updated based on a gradient criteria, regarding the past two values of the output power. 3 Regimes are established, concerning ramp events: ramp-up, ramp-down and no-ramp regime. In order to assess the skillness of the proposed RS-ANN model, a single

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

    Science.gov (United States)

    Venkiteswaran, C. N.; 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.

  17. 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.

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

    Directory of Open Access Journals (Sweden)

    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.

  19. 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.

  20. 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.

  1. Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

    Science.gov (United States)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal, Asiah, Nur; Shafii, M. Ali

    2010-12-01

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to 70%. Two dimensional full core multi groups diffusion calculations was performed by CITATION code. Group constant preparations are performed by using SRAC code system with JENDL-3.2 nuclear data library. In this design the reactor cores with cylindrical cell two dimensional R-Z core models are subdivided into several parts with the same volume in the axial directions. The placement of fuel in core arranged so that the result of plutonium from natural uranium can be utilized optimally for 10 years reactor operation. Modified CANDLE burn-up was established successfully in a core radial width 1.4 m. Total thermal power output for reference core is 550 MW. Study on the effect of fuel to coolant ratio shows that effective multiplication factor (keff) is in almost linear relations with the change of the fuel volume to coolant ratio.

  2. Stochastic modeling of the dynamics of incident-induced lane traffic states for incident-responsive local ramp control

    Science.gov (United States)

    Sheu, Jiuh-Biing

    2007-12-01

    Incident-induced traffic congestion has been recognized as a critical issue to solve in the development of advanced freeway incident management systems. This paper investigates the applicability of a stochastic optimal control approach to real-time incident-responsive local ramp control on freeways. The architecture of the proposed ramp control system embeds two primary functions including (1) real-time estimation of incident-induced lane traffic states and (2) dynamic prediction of ramp-metering rates in response to the changes of incident impacts. To accomplish the above two goals, a discrete-time nonlinear stochastic optimal control model is proposed, followed by the development of a recursive prediction algorithm. Based on the simulation data, the numerical results of model tests indicate that the proposed method permits relieving incident impacts particularly under low-volume and medium-volume conditions, relative to high-volume lane-blocking conditions. Particularly, the incident-induced queue lengths can be improved by 50.1% and 67.9%, compared to the existing ramp control and control-free strategies, respectively.

  3. 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)

  4. Early Callovian ingression in southwestern Gondwana. Palaeoenvironmental evolution of the carbonate ramp (Calabozo Formation) in southwestern Mendoza, Neuquen basin, Argentina

    Science.gov (United States)

    Armella, Claudia; Cabaleri, Nora G.; Cagnoni, Mariana C.; Panarello, Héctor O.

    2013-08-01

    The carbonatic sequence of the Calabozo Formation (Lower Callovian) developed in southwestern Gondwana, within the northern area of the Neuquén basin, and is widespread in thin isolated outcrops in southwestern Mendoza province, Argentina. This paper describes the facies, microfacies and geochemical-isotopic analysis carried out in five studied localities, which allowed to define the paleoenvironmental conditions of a homoclinal shallow ramp model, highly influenced by sea level fluctuations, where outer, mid and inner ramp subenvironments were identified. The outer ramp subenvironment was only recognized in the south of the depocenter and is characterized by proximal outer ramp facies with shale levels and interbedded mudstone and packstone layers. The mid ramp subenvironment is formed by low energy facies (wackestone) affected by storms (packstones, grainstones and floatstones). The inner ramp subenvironment is the most predominant and is characterized by tidal flat facies (wackestones, packstones and grainstones) over which a complex of shoals (grainstones and packstones) dissected by tidal channels (packstone, grainstones and floatstones) developed. In the north area, protected environment facies were recorded (bioturbated wackestones and packstones). The vertical distribution of facies indicates that the paleoenvironmental evolution of the Calabozo Formation results from a highstand stage in the depocenter, culminating in a supratidal environment, with stromatolitic levels interbedded with anhydrite originated under restricted water circulation conditions due to a progressive isolation of the basin. δ13C and δ18O values of the carbonates of the Calabozo Formation suggest an isotopic signature influenced by local palaeoenvironmental parameters and diagenetic overprints. The δ13C and δ18O oscillations between the carbonates of the different studied sections are related with lateral facies variations within the carbonate ramp accompanied with dissimilar

  5. Reinforcement Learning for Ramp Control: An Analysis of Learning Parameters

    Directory of Open Access Journals (Sweden)

    Chao Lu

    2016-08-01

    Full Text Available Reinforcement Learning (RL has been proposed to deal with ramp control problems under dynamic traffic conditions; however, there is a lack of sufficient research on the behaviour and impacts of different learning parameters. This paper describes a ramp control agent based on the RL mechanism and thoroughly analyzed the influence of three learning parameters; namely, learning rate, discount rate and action selection parameter on the algorithm performance. Two indices for the learning speed and convergence stability were used to measure the algorithm performance, based on which a series of simulation-based experiments were designed and conducted by using a macroscopic traffic flow model. Simulation results showed that, compared with the discount rate, the learning rate and action selection parameter made more remarkable impacts on the algorithm performance. Based on the analysis, some suggestionsabout how to select suitable parameter values that can achieve a superior performance were provided.

  6. Cumulative Interarrival Time Distributions of Freeway Entrance Ramp Traffic for Traffic Simulations

    Directory of Open Access Journals (Sweden)

    Erdinç Öner

    2013-02-01

    Full Text Available Cumulative interarrival time (IAT distributions for signalized and non-signalized freeway entrance ramps were developed to be used in digital computer traffic simulation models. The data from four different non-signalized entrance ramps (three ramps with a single lane, one ramp with two lanes and two different signalized entrance ramps (both with a single lane were used for developing the cumulative IAT distributions. The cumulative IAT distributions for the signalized and non-signalized entrance ramps were compared with each other and with the cumulative IAT distributions of the lanes for freeways. The comparative results showed that the cumulative IAT distributions for non-signalized entrance ramps are very close to the leftmost lane of a 3-lane freeway where the maximum absolute difference between the cumulative IAT distribution of the leftmost lane of a 3-lane freeway and the entrance ramps cumulative IAT distribution was 3%. The cumulative IAT distribution for the signalized entrance ramps was found to be different from the non-signalized entrance ramp cumulative IAT distribution. The approximated cumulative IAT distributions for signalized and non-signalized entrance ramp traffic for any hourly traffic volume from a few vehicles/hour up to 2,500 vehicles/hour can be obtained at http://www.ohio.edu/orite/research/uitds.cfm.

  7. Radar echo processing with partitioned de-ramp

    Science.gov (United States)

    Dubbert, Dale F.; Tise, Bertice L.

    2013-03-19

    The spurious-free dynamic range of a wideband radar system is increased by apportioning de-ramp processing across analog and digital processing domains. A chirp rate offset is applied between the received waveform and the reference waveform that is used for downconversion to the intermediate frequency (IF) range. The chirp rate offset results in a residual chirp in the IF signal prior to digitization. After digitization, the residual IF chirp is removed with digital signal processing.

  8. Adiabaticity of the ramping process of an ac dipole

    Directory of Open Access Journals (Sweden)

    R. Tomás

    2005-02-01

    Full Text Available ac dipoles in accelerators are used to excite coherent betatron oscillations at a drive frequency close to the tune. If the excitation amplitude is slowly increased to the desired value and slowly decreased back to zero there is no significant emittance growth. The aim of this article is to study the adiabaticity of the ramping process of an ac dipole as a function of the different parameters involved.

  9. Kinematics investigations of cylinders rolling down a ramp using tracker

    Science.gov (United States)

    Prima, Eka Cahya; Mawaddah, Menurseto; Winarno, Nanang; Sriwulan, Wiwin

    2016-02-01

    Nowadays, students' exploration as well as students' interaction in the application stage of learning cycle can be improved by directly model real-world objects based on Newton's Law using Open Source Physics (OSP) computer-modeling tools. In a case of studying an object rolling down a ramp, a traditional experiment method commonly uses a ticker tape sliding through a ticker timer. However, some kinematics parameters such as the instantaneous acceleration and the instantaneous speed of object cannot be investigated directly. By using the Tracker video analysis method, all kinematics parameters of cylinders rolling down a ramp can be investigated by direct visual inspection. The result shows that (1) there are no relations of cylinders' mass as well as cylinders' radius towards their kinetics parameters. (2) Excluding acceleration data, the speed and position as function of time follow the theory. (3) The acceleration data are in the random order, but their trend-lines closely fit the theory with 0.15% error. (4) The decrease of acceleration implicitly occurs due to the air friction acting on the cylinder during rolling down. (5) The cylinder's inertial moment constant has been obtained experimentally with 3.00% error. (6) The ramp angle linearly influences the cylinders' acceleration with 2.36% error. This research implied that the program can be further applied to physics educational purposes.

  10. Review of Wind Energy Forecasting Methods for Modeling Ramping Events

    Energy Technology Data Exchange (ETDEWEB)

    Wharton, S; Lundquist, J K; Marjanovic, N; Williams, J L; Rhodes, M; Chow, T K; Maxwell, R

    2011-03-28

    Tall onshore wind turbines, with hub heights between 80 m and 100 m, can extract large amounts of energy from the atmosphere since they generally encounter higher wind speeds, but they face challenges given the complexity of boundary layer flows. This complexity of the lowest layers of the atmosphere, where wind turbines reside, has made conventional modeling efforts less than ideal. To meet the nation's goal of increasing wind power into the U.S. electrical grid, the accuracy of wind power forecasts must be improved. In this report, the Lawrence Livermore National Laboratory, in collaboration with the University of Colorado at Boulder, University of California at Berkeley, and Colorado School of Mines, evaluates innovative approaches to forecasting sudden changes in wind speed or 'ramping events' at an onshore, multimegawatt wind farm. The forecast simulations are compared to observations of wind speed and direction from tall meteorological towers and a remote-sensing Sound Detection and Ranging (SODAR) instrument. Ramping events, i.e., sudden increases or decreases in wind speed and hence, power generated by a turbine, are especially problematic for wind farm operators. Sudden changes in wind speed or direction can lead to large power generation differences across a wind farm and are very difficult to predict with current forecasting tools. Here, we quantify the ability of three models, mesoscale WRF, WRF-LES, and PF.WRF, which vary in sophistication and required user expertise, to predict three ramping events at a North American wind farm.

  11. 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.

  12. Development of ramp-flat structures during Aegean extension

    Science.gov (United States)

    Brun, Jean-Pierre; Sokoutis, Dimitrios

    2014-05-01

    Low-angle extensional shear is frequently observed in the Aegean metamorphic rocks. This deformation is commonly interpreted as being related to detachment at crustal scale, yet it often corresponds to ramp-flat extensional systems that, at many places, control the deposition of Neogene sedimentary basins. From a mechanical point of view, the development of a ramp-flat structure requires the presence of weak layers that can be activated as décollement between stronger rocks units. In the Aegean, the décollement generally develops within the upper brittle crust (i.e. with temperatures lower than about 400°C) that consists in recently exhumed metamorphic rocks. The process by which, these layers become weak enough to form efficient décollements in extension is somewhat intriguing and not well understood. In this contribution we examine the particular case of ramp-flat structures of the Southern Rhodope Core Complex that controlled the deposition of late Miocene to Pleistocene sediments in continental and marine basins. Field evidence is used to argue that the décollement corresponds to marble layers that separate orthogneisses at 2-3 km depth within an upper brittle crust whose thickness is around 5 km. Field observation and stable isotope measurements suggest that the ramp-flat structure observed on the island of Thassos occurred in a marble unit rich in fluids at a temperature of around 200°C. Using laboratory experiments, we explore the geometry of extensional structures (fault systems, rollovers, piggy-back basins…) that can develop at crustal-scale as a function of: i) décollement depth and dip, ii) number of décollements, and iii) strength contrast, between the décollement and overlying strong units. The results are compared with the situation observed in the Southern Rhodope Core Complex. We are convinced that the principles of ramp-flat extension discussed here have a strong potential of application in many other orogenic domains affected by large

  13. XS-1 on ramp with B-29 mothership

    Science.gov (United States)

    1949-01-01

    XS-1 on the ramp with the B-29 mothership in 1949. This is the second XS-1 built; it later was converted into the X-1E. Unlike the XS-1-1, which was flown by the Air Force, the XS-1-2 was flown mostly by Bell and NACA pilots. It gathered much more research data than the more famous XS-1-1, known as 'Glamorous Glennis.' The first of the rocket-powered research aircraft, the X-1 (originally designated the XS-1), was a bullet-shaped airplane that was built by the Bell Aircraft Company for the US Air Force and the NACA. The mission of the X-1 was to investigate the transonic speed range (speeds from just below to just above the speed of sound) and, if possible, to break the 'sound barrier.' The first of the three X-1s was glide-tested at Pinecastle Army Air Field, FL, in early 1946. The first powered flight of the X-1 was made on Dec. 9, 1946, at Edwards Air Force Base with Chalmers Goodlin, a Bell test pilot, at the controls. On Oct. 14, 1947, with USAF Captain Charles 'Chuck' Yeager as pilot, the aircraft flew faster than the speed of sound for the first time. Captain Yeager ignited the four-chambered XLR-11 rocket engines after the B-29 air-launched it from under the bomb bay of a B-29 at 21,000 feet. The 6,000-pound thrust ethyl alcohol/liquid oxygen burning rockets, built by Reaction Motors, Inc., pushed the aircraft up to a speed of 700 miles per hour in level flight. Captain Yeager was also the pilot when the X-1 reached its maximum speed, 957 miles per hour. Another USAF pilot. Lt. Col. Frank Everest, Jr., was credited with taking the X-1 to its maximum altitude of 71,902 feet. Eighteen pilots in all flew the X-1s. The number three plane was destroyed in a fire before ever making any powered flights. A single-place monoplane, the X-1 was 30 feet, 11 inches long; 10 feet, 10 inches high; and had a wingspan of 29 feet. It weighed 6,784 pounds and carried 6,250 pounds of fuel. It had a flush cockpit with a side entrance and no ejection seat.

  14. Neutronic Study of Burnup, Radiotoxicity, Decay Heat and Basic Safety Parameters of Mono-Recycling of Americium in French Pressurised Water Reactors

    Directory of Open Access Journals (Sweden)

    Robert Bright Mawuko Sogbadji

    2017-03-01

    Full Text Available The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu. This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am in present French pressurised water reactors (PWR. These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a ‘waiting strategy’ for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and

  15. Factors controlling the sedimentary evolution of the Kimmeridgian ramp in the north Iberian Basin (NE Spain

    Directory of Open Access Journals (Sweden)

    Bádenas, B.

    1994-04-01

    Full Text Available The aim of this paper is to summarize the present knowledge reached by the authors on the carbonate ramp which developed in the iberian basin during Kimmeridgian times. Our results were obtained from a combined field analysis and computer modelling carried out in the north Iberian Chain (NE Spain. Extensive field analysis in the Ricia area (Zaragoza, NE Spain, resulted in a detailed mapping of the transition from inner to outerramp facies on this carbonate rampo Three facies belts may be distinguished in this rampo The outer ramp facies consists of marls and mudstones rhythmic facies. The inner ramp facies, located aboye fair-weather wave base, are dominated by coral patch reef growing. The middle ramp facies are represented by marls and micrites bearing skeletal and oolitic tempestite levels which sharply grade into high-amplitude o'olitic sandwave. Factors such as resedimentation by storms, carbonate production and relative variation of sea level acting in the Kimmeridgian ramp are also quantiphied and discussed. Most of the mud accumulated in outer-ramp areas was produced in the coral «carbonate factory» located in inner areas. Off-shore resedimentation by storm was the main agent of basinward transport of this mudo The deduced accommodation curve consists of three elements: a linear rise which satisfactorily matches the normal subsidence figures observed in intracratonic basins; a third-order cycle, that may have a regional cause and higher order cycles in the Milanckovich band, that may be eustatic in origin.La sedimentación en la cuenca ibérica septentrional durante el Kimmeridgiense tuvo lugar en una extensa rampa carbonatada de bajo ángulo. Las facies de rampa externa, acumuladas por debajo del nivel de base del oleaje debido a tormentas (i.e., C. 50 to 80 m de profundidad, están formadas por una ritmita de margas y calizas (i.e., Fm Loriguilla. Las facies de rampa interna, localizadas por encima del nivel de base del oleaje de

  16. Use of burnup credit in criticality evaluation for spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Chon, Je Keun; Kim, Jae Chun; Koh, Duck Joon; Kim Byung Tae [Nuclear Environment Technology Institute, Korea Electric Power Corporation, Taejon (Korea, Republic of)

    1999-07-01

    Boraflex is a polymer based material which is used as matrix to contain a neutron absorber material, boron carbide. In a typical spent fuel pool the irradiated Boraflex has been known as a significant source of silica. Since 1996, it was reported that elevated silica levels were measured in the Ulchin Unit 2 spent fuel pool water. Therefore, the Ulchin Unit 2 spent fuel storage racks were needed to be reanalyzed to allow storage of fuel assemblies with normal enrichments up to 5.0w/o U-235 in all storage cell locations using credit for burnup. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels. In region 2, the calculations were performed by assuming in an infinite radial array of storage cells. No credit is taken for axial or radial neutron leakage. The water in the spent fuel storage pool was assumed to be pure. In the evaluation of the Ulchin Unit 2 spent fuel storage pool, criticality analyses were performed with the CASMO-3 code. A reactivity uncertainty in the fuel depletion calculations was combined with other calculational uncertainty. The manufacturing tolerances were considered, as well. From the calculation, the acceptable burnup domain in region 2 of the spent fuel storage pool. where the curve identifies conditions of equal reactivity for various initial enrichments between 1.6w/o and 5.0w/o, was evaluated. In region 2, the maximum k{sub e}ff including all uncertainties, is 0.94648 for the enrichment-burnup combination from loading curve. (author)

  17. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  18. Spent fuel dissolution rates as a function of burnup and water chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  19. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  20. 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.

  1. Getting to the On-ramp of the Information Superhighway

    Science.gov (United States)

    1996-01-01

    manager may elect to automatically forward all incoming e- mail from selected senders to his deputy. It is not necessary to have an AMH installed if users...establish a process to quickly identify the important and rou- tine messages. Most e- mail systems have an inbox which sorts unread e- mail and...messages FYI: for all �unofficial� For Your In- 27 Getting to the On-Ramp formation, or optional messages E- mail should also be � sender friendly.� Some

  2. Online Analysis of Wind and Solar Part I: Ramping Tool

    Energy Technology Data Exchange (ETDEWEB)

    Etingov, Pavel V.; Ma, Jian; Makarov, Yuri V.; Subbarao, Krishnappa

    2012-01-31

    To facilitate wider penetration of renewable resources without compromising system reliability concerns arising from the lack of predictability of intermittent renewable resources, a tool for use by California Independent System Operator (CAISO) power grid operators was developed by Pacific Northwest National Laboratory (PNNL) in conjunction with CAISO with funding from California Energy Commission. This tool predicts and displays additional capacity and ramping requirements caused by uncertainties in forecasts of loads and renewable generation. The tool is currently operational in the CAISO operations center. This is one of two final reports on the project.

  3. Hydrogeology of the unsaturated zone, North Ramp area of the Exploratory Studies Facility, Yucca Mountain, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, J.P.; Kwicklis, E.M.; Gillies, D.C. [eds.

    1999-03-01

    Yucca Mountain, in southern Nevada, is being investigated by the US Department of Energy as a potential site for a repository for high-level radioactive waste. This report documents the results of surface-based geologic, pneumatic, hydrologic, and geochemical studies conducted during 1992 to 1996 by the US Geological Survey in the vicinity of the North Ramp of the Exploratory Studies Facility (ESF) that are pertinent to understanding multiphase fluid flow within the deep unsaturated zone. Detailed stratigraphic and structural characteristics of the study area provided the hydrogeologic framework for these investigations. Shallow infiltration is not discussed in detail in this report because the focus in on three major aspects of the deep unsaturated-zone system: geologic framework, the gaseous-phase system, and the aqueous-phase system. However, because the relation between shallow infiltration and deep percolation is important to an overall understanding of the unsaturated-zone flow system, a summary of infiltration studies conducted to date at Yucca Mountain is provided in the section titled Shallow Infiltration. This report describes results of several Site Characterization Plan studies that were ongoing at the time excavation of the ESF North Ramp began and that continued as excavation proceeded.

  4. Void effect analysis of Pb-208 of fast reactors with modified CANDLE burn-up scheme

    Science.gov (United States)

    Widiawati, Nina; Su'ud, Zaki

    2015-09-01

    Void effect analysis of Pb-208 as coolant of fast reactors with modified candle burn-up scheme has been conducted. Lead cooled fast reactor (LFR) is one of the fourth-generation reactor designs. The reactor is designed with a thermal power output of 500 MWt. Modified CANDLE burn-up scheme allows the reactor to have long life operation by supplying only natural uranium as fuel cycle input. This scheme introducing discrete region, the fuel is initially put in region 1, after one cycle of 10 years of burn up it is shifted to region 2 and region 1 is filled by fresh natural uranium fuel. The reactor is designed for 100 years with 10 regions arranged axially. The results of neutronic calculation showed that the void coefficients ranged from -0.6695443 % at BOC to -0.5273626 % at EOC for 500 MWt reactor. The void coefficients of Pb-208 more negative than Pb-nat. The results showed that the reactors with Pb-208 coolant have better level of safety than Pb-nat.

  5. Tritium Burn-up Depth and Tritium Break-Even Time

    Institute of Scientific and Technical Information of China (English)

    LI Cheng-Yue; DENG Bai-Quan; HUANG Jin-Hua; YAN Jian-Cheng

    2006-01-01

    @@ Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than theleast required amount of tritium storage to start up three of FEB-like fusion reactors.

  6. 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.

  7. 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.

  8. Comparison of neutron cross sections for selected fission products and isotopic composition analyses with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Gil, C. S.; Kim, J. D.; Jang, J. H.; Lee, Y. D. [KAERI, Taejon (Korea)

    2003-10-01

    The neutron absorption cross sections for 18 fission products evaluated within the framework of the KAERI-BNL international collaboration have been compared with the ENDF/B-VI release 7. Also, the influence of the new evaluations on isotopic compositions of the fission products as a function of burnup has been analyzed through the OECD/NEA burnup credit criticality benchmarks (Phase 1B) and the LWR/Pu recycling benchmarks. These calculations were performed by WIMSD-5B with the 69 group libraries prepared from three evaluated nuclear data libraries: ENDF/B-VI.7, ENDF/B-VI.8 including new evaluations in resonance region covering thermal region, and ENDF/B-VII expected including those in upper resonance region up to 20 MeV. For Xe-131, the composition calculated with ENDF/B-VI.8 shows maximum difference of 4.78% compared to ENDF/B-VI.7. However, the isotopic compositions of all fission products calculated with ENDF/B-VII shows no differences compared to ENDF/B-VI.7.

  9. Oxygen uptake kinetics during incremental- and decremental-ramp cycle ergometry.

    Science.gov (United States)

    Ozyener, Fadil; Rossiter, Harry B; Ward, Susan A; Whipp, Brian J

    2011-01-01

    The pulmonary oxygen uptake (VO2) response to incremental-ramp cycle ergometry typically demonstrates lagged-linear first-order kinetics with a slope of ~10-11 ml·min(-1)·W(-1), both above and below the lactate threshold (θL), i.e. there is no discernible VO2 slow component (or "excess" VO2) above θL. We were interested in determining whether a reverse ramp profile would yield the same response dynamics. Ten healthy males performed a maximum incremental -ramp (15-30 W·min(-1), depending on fitness). On another day, the work rate (WR) was increased abruptly to the incremental maximum and then decremented at the same rate of 15-30 W.min(-1) (step-decremental ramp). Five subjects also performed a sub-maximal ramp-decremental test from 90% of θL. VO2 was determined breath-by-breath from continuous monitoring of respired volumes (turbine) and gas concentrations (mass spectrometer). The incremental-ramp VO2-WR slope was 10.3 ± 0.7 ml·min(-1)·W(-1), whereas that of the descending limb of the decremental ramp was 14.2 ± 1.1 ml·min(-1)·W(-1) (p incremental-ramp. This suggests that the VO2 response in the supra-θL domain of incremental-ramp exercise manifest not actual, but pseudo, first-order kinetics. Key pointsThe slope of the decremental-ramp response is appreciably greater than that of the incremental.The response dynamics in supra-θL domain of the incremental-ramp appear not to manifest actual first-order kinetics.The mechanisms underlying the different dynamic response behaviour for incremental and decremental ramps are presently unclear.

  10. AN ACCURATE MODELING OF DELAY AND SLEW METRICS FOR ON-CHIP VLSI RC INTERCONNECTS FOR RAMP INPUTS USING BURR’S DISTRIBUTION FUNCTION

    Directory of Open Access Journals (Sweden)

    Rajib Kar

    2010-09-01

    Full Text Available This work presents an accurate and efficient model to compute the delay and slew metric of on-chip interconnect of high speed CMOS circuits foe ramp input. Our metric assumption is based on the Burr’s Distribution function. The Burr’s distribution is used to characterize the normalized homogeneous portion of the step response. We used the PERI (Probability distribution function Extension for Ramp Inputs technique that extends delay metrics and slew metric for step inputs to the more general and realistic non-step inputs. The accuracy of our models is justified with the results compared with that of SPICE simulations.

  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. New energy ramping control system in the Pohang Light Source Storage Ring

    CERN Document Server

    Kim, E S; Seo, J H; Nam, S H

    2003-01-01

    A new energy ramping control system is developed to increase electron beam energy from 2.0 GeV to 2.5 GeV in the Pohang Light Source storage ring. The control system shows faster energy ramping and more stable beam characteristics than the previous energy ramping control system that consists of an operator interface workstation, subsystem control computer and machine interface units. The previous system controls current settings of the magnet power supplies in multiple steps with different energy increment rate during the energy ramping. In order to improve synchronization of the magnet power supply systems and beam stabilities during the energy ramping, we developed the energy ramping controller hardware that is controlled using a network of optical fibers. Thus, the new ramping system is able to control more synchronously the magnet power supplies with a constant energy increment rate during the energy ramping. With this reliable synchronization, it is also possible to perform the energy de-ramping step fro...

  13. Parametric investigation of single-expansion-ramp nozzles at Mach numbers from 0.60 to 1.20

    Science.gov (United States)

    Capone, Francis J.; Re, Richard J.; Bare, E. Ann

    1992-01-01

    An investigation was conducted in the Langley 16-Foot Transonic Tunnel to determine the effects of varying six nozzle geometric parameters on the internal and aeropropulsive performance characteristics of single-expansion-ramp nozzles. This investigation was conducted at Mach numbers from 0.60 to 1.20, nozzle pressure ratios from 1.5 to 12, and angles of attack of 0 deg +/- 6 deg. Maximum aeropropulsive performance at a particular Mach number was highly dependent on the operating nozzle pressure ratio. For example, as the nozzle upper ramp length or angle increased, some nozzles had higher performance at a Mach number of 0.90 because of the nozzle design pressure was the same as the operating pressure ratio. Thus, selection of the various nozzle geometric parameters should be based on the mission requirements of the aircraft. A combination of large upper ramp and large lower flap boattail angles produced greater nozzle drag coefficients at Mach number greater than 0.80, primarily from shock-induced separation on the lower flap of the nozzle. A static conditions, the convergent nozzle had high and nearly constant values of resultant thrust ratio over the entire range of nozzle pressure ratios tested. However, these nozzles had much lower aeropropulsive performance than the convergent-divergent nozzle at Mach number greater than 0.60.

  14. Tailored ramp wave generation in gas gun experiments

    Directory of Open Access Journals (Sweden)

    Cotton Matthew

    2015-01-01

    Full Text Available Gas guns are traditionally used as platforms to introduce a planar shock wave to a material using plate impact methods, generating states on the Hugoniot. The ability to deliver a ramp wave to a target during a gas gun experiment enables access to different regions of the equation-of-state surface, making it a valuable technique for characterising material behaviour. Previous techniques have relied on the use of multi-material impactors to generate a density gradient, which can be complex to manufacture. In this paper we describe the use of an additively manufactured steel component consisting of an array of tapered spikes which can deliver a ramp wave over ∼ 2 μs. The ability to tailor the input wave by varying the component design is discussed, an approach which makes use of the design freedom offered by additive manufacturing techniques to rapidly iterate the spike profile. Results from gas gun experiments are presented to evaluate the technique, and compared with 3D hydrodynamic simulations.

  15. Changes of hydrodynamic parameters on mountain stream bed within the block ramp influence and possibility of their use for integrated river management

    Science.gov (United States)

    Radecki-Pawlik, Artur; Plesiński, Karol

    2016-04-01

    In modern river management practices and philosophy one can notice coming more into use ecological friendly hydraulic structures. Those, which are especially needed for river training works, as far as expectation of Water Framework Directive is concerned, are block ramps which are hydraulic structures working similar to riffles known very well from fluvial geomorphology studies and are natural features in streams and rivers. What is important well designed block ramps do not stop fish and invertebrates against migrating, provide natural and esthetical view being built within the river channel, still working as hydraulic engineering structures and might be used in river management in different river ecosystems. The main aim of the research was to describe changes of values of hydrodynamics parameters upstream and downstream of the block ramps and to find out their influence on hydrodynamics of the stream. The study was undertaken on the Porębianka River in the Gorce Mountains, Polish Carpathians. Observed hydrodynamic parameters within the reach of the block ramps depend on the location of measuring point and the influence of individual part of the structure. We concluded that: 1. Hydrodynamic parameters close to block ramps depend on the location of the measurement points in relation to particular elements of the structure; 2. The highest value of velocities don't cause the highest force values, which acting on the bed of the watercourse, because they are rather related to the water level of the channel; 3. The values of mean velocities, shear velocities and shear stresses were similar upstream and downstream the block ramps, which means that the structures stabilize the river bed. This study was performed within the scope of the Science Activity money from Ministry of High Education and Young Scientist's Activity Money of Department of Hydraulics Engineering and Geotechnique, University of Agriculture, Cracow, Poland

  16. Effect of station-specific alerting and ramp-up tones on firefighters' alarm time heart rates.

    Science.gov (United States)

    MacNeal, James J; Cone, David C; Wistrom, Christopher L

    2016-11-01

    A number of long-term health effects are suffered by emergency responders, some influenced by psychological stress and fatigue. This study explored if stress and fatigue can be reduced by changing the method by which firefighters are alerted to emergency responses. Over several months, the method by which responders at a fire department were alerted was altered. Firefighter heart rates were measured first with standard alerting as a control (phase 1: all stations alerted simultaneously, with high-volume tones). The department then implemented station-specific (phase 2) and gradual volume ramp-up (phase 3) tone alerting, and heart rate increases were compared. The Fatigue Severity Score was used to examine firefighter fatigue, and the department administered a follow-up survey on personnel preferences. Individual heart rate increases (Δbpm) ranged from 2-48 bpm. Median increases were 7 bpm (IQR 4-11 bpm) during phase 1 (72.2% of alarms Δbpmalarms Δbpmalarms Δbpmalarms in phase 3 resulted in increases of Fatigue Severity Scale showed little variability: median scores 7 in phase 1, 8 in phase 2, and 7 in phase 3. Firefighters reported a strong preference for the "ramp-up" tones, and were roughly evenly divided between preferring alerting all stations simultaneously 24/7 (40% rating this 4 or 5 on a five-point Likert scale), station-specific alerting 24/7 (47.5%), or all stations during the day but station-specific at night (40%). Ramp-up tones were perceived as the best method to reduce stress during the day and overnight. Small but significant decreases in the amount of tachycardic response to station alerting are associated with simple alterations in alerting methods. Station-specific and ramp-up tones improve perceived working conditions for emergency responders.

  17. Study of electro-optic effect in asymmetrically ramped AlInGaAs multiple quantum well structures

    Energy Technology Data Exchange (ETDEWEB)

    Sadiq, Muhammad Usman; Peters, Frank H.; Corbett, Brian [Tyndall National Institute, Lee Maltings, Cork (Ireland); Department of Physics, University College Cork, Cork (Ireland); O' Callaghan, James; Roycroft, Brendan; Thomas, Kevin; Pelucchi, Emanuele [Tyndall National Institute, Lee Maltings, Cork (Ireland)

    2016-04-15

    We investigate the electro-optic properties of two oppositely ramped asymmetric quantum well structures in the AlInGaAs material system. The grading of the bandgap in the quantum wells has been achieved by changing the ratio of Al to Ga in the quaternary alloy during the epitaxial growth. The surface normal photo-response and the Fabry-Perot fringe shift in straight waveguides are compared for both structures as a function of applied voltage at 1550 nm for TE-polarized light. The measurements show a change in the refractive index due to a red shift of the excitonic resonances due to the quantum-confined Stark effect. The 10 quantum well structure with a ramp up of the bandgap in the growth direction leads to the figure of merit of the voltage for a π phase shift, V{sub π} by length, L, V{sub π} x L, of 6 as compared to 7 V . mm in the structure with a ramp in opposite direction. Further investigations show that the reduction in V{sub π} is due to increased absorption at high reverse bias which induces a non-linear phase change. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  18. Atomic Force Microscopy Thermally-Assisted Microsampling with Atmospheric Pressure Temperature Ramped Thermal Desorption/Ionization-Mass Spectrometry Analysis.

    Science.gov (United States)

    Hoffmann, William D; Kertesz, Vilmos; Srijanto, Bernadeta R; Van Berkel, Gary J

    2017-02-20

    The use of atomic force microscopy controlled nanothermal analysis probes for reproducible spatially resolved thermally assisted sampling of micrometer-sized areas (ca. 11 × 17 μm wide × 2.4 μm deep) from relatively low number-average molecular weight (Mn mass spectrometric analysis. The procedure and mechanism for material pickup, the sampling reproducibility and sampling size are discussed, and the oligomer distribution information available from slow temperature ramps versus ballistic temperature jumps is presented. For the Mn = 970 P2VP, the Mn and polydispersity index determined from the mass spectrometric data were in line with both the label values from the sample supplier and the value calculated from the simple infusion of a solution of polymer into the commercial atmospheric pressure chemical ionization source on this mass spectrometer. With a P2VP sample of higher Mn (Mn = 2070 and 2970), intact oligomers were still observed (as high as m/z 2793 corresponding to the 26-mer), but a significant abundance of thermolysis products were also observed. In addition, the capability for confident identification of the individual oligomers by slowly ramping the probe temperature and collecting data-dependent tandem mass spectra was also demonstrated. The material type limits to the current sampling and analysis approach as well as possible improvements in nanothermal analysis probe design to enable smaller area sampling and to enable controlled temperature ramps beyond the present upper limit of about 415 °C are also discussed.

  19. Analysis of the EDIPO Temperature Margin During Current Ramp-Up

    CERN Document Server

    Marinucci, C; Calvi, Marco; Marinucci, Claudio; Cau, Francesca; Bottura, Luca

    2010-01-01

    The European dipole (EDIPO), currently under construction, will provide background magnetic fields of up to 12.5 T for tests of ITER high-current superconducting cables. The EDIPO winding consists of 7 x 2 double layers of Nb3SN cable-in-conduit conductors with forced flow cooling of supercritical helium. The performance limits of EDIPO during current ramp-up are analyzed analysed with the CryoSoft suite of codes, recently integrated into a customizable and flexible environment for the analysis of thermal hydraulic and electrical transients in superconducting magnetic systems. The simultaneous analysis of the cryogenic system and all 14 double layers shows that under all charging conditions the EDIPO temperature margin remains sufficiently high.

  20. Conceptual design study on an upgraded future Monju core (2). Core concept with extended refueling interval and increased fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kinjo, Hidehito; Ishibashi, Jun-ichi; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, International Cooperation and Technology Development Center, Tsuruga, Fukui (Japan); Kageyama, Takeshi [Nuclear Energy System Inc., Tokyo (Japan)

    2003-03-01

    A conceptual design study has been performed at the International Cooperation and Technology Development Center to investigate the feasibility of upgraded future Monju cores with extended refueling intervals of 365efpd/cycle and increased fuel burnup of 150 GWd/t. The goal of this study is to demonstrate the possible contribution of Monju to the improved economy and to efficient utilization, as one of the major facilities for fast neutron irradiation. Two design measures have been mainly taken to improve the core fuel burnup and reactivity control characteristics for the extended operating cycle length of 1 year: (1) The driver fuel pin specification with both increased pin diameter of 7.7mm and increased active core height of about 100cm has been chosen to reduce the burnup reactivity swing, (2) The absorber control rod specification has also been changed to enhance the control rod reactivity worth by increasing {sup 10}B-enrichment and absorber length, and to adequately secure the shutdown reactivity margin. The major core characteristics have been evaluated on the core power distribution, safety parameters such as sodium void reactivity and Doppler effect, thermal hydraulics and reactivity control characteristics. The results show that this core could achieve the targeted core performances of 1-year operating cycle as well as 150GWd/t discharged burnup, without causing any significant drawback on the core characteristics and safety aspects. The upgraded core concepts have, therefore, been confirmed as feasible. (author)

  1. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  2. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  3. Oolitic sandbody depositional models and geometries, Mississippian of southwest Britain: implications for petroleum exploration in carbonate ramp settings

    Science.gov (United States)

    Burchette, Trevor P.; Paul Wright, V.; Faulkner, Tom J.

    1990-07-01

    A 1000 m thick early Mississippian carbonate supersequence, the "Carboniferous Limestone" of southwest Britain, consists of three third-order depositional sequences. These comprise parasequences in various configurations, and the whole forms a carbonate ramp stack. Within this framework five major oolitic carbonate sandbodies developed: (a) Castell Coch Limestone, (b) Stowe Oolite, (c) Brofiscin Oolite, (d) Gully Oolite, and (e) High Tor Limestone. The depositional regime was storm- and wave-dominated throughout and the major sandbodies represent a range of progradational carbonate beaches, barriers and detached subtidal shoals. Analysis of the three-dimensional shapes and distribution of these five examples shows that they evolved to produce three major carbonate sandbody geometries: (a) strings, (b) sheets, and (c) wedges. These geometries are characterised using the five field examples and offered as a template which may assist in the exploration and reservoir modelling of petroleum-rich high-energy ramp systems. Progradation, for up to 40 km, of barrier islands (Stowe Oolite) and beach-ridge plains (Gully Oolite Formation) generated strings, and "thick" sheets individually up to 10-20 m thick. "Thin" shoreface-retreat carbonate packstone/grainstone sheets up to 5 m thick (High Tor limestone) developed during transgressions as veneers across flooding surfaces. These are comparable with sheet sands developed in siliciclastic shelf depositional systems. Progradation, for up to 30 km, and vertical aggradation of shoreline-detached oolite shoals (Castell Coch limestone, Brofiscin Oolite), generated basinwards-expanding or thinning wedges up to 30 m thick. Tectonically controlled stacking of strandplain sheets produced a composite carbonate sandbody up to 80 m thick (Gully Oolite). The intrinsic (sedimentary) and extrinsic (eustacy, tectonism, climate) factors which controlled these sandbody geometries are addressed. Establishing the positions of the sandbodies

  4. Thermal ramp tritium release in COBRA-1A2 C03 beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, D.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Tritium release kinetics, using the method of thermal ramp heating at three linear ramp rates, were measured on the COBRA-1A2 C03 1-mm beryllium pebbles. This report includes a brief discussion of the test, and the test data in graph format.

  5. Effects of compression and expansion ramp fuel injector configuration on scramjet combustion and heat transfer

    Science.gov (United States)

    Stouffer, Scott D.; Baker, N. R.; Capriotti, D. P.; Northam, G. B.

    1993-01-01

    A scramjet combustor with four wall-ramp injectors containing Mach-1.7 fuel jets in the base of the ramps was investigated experimentally. During the test program, two swept ramp injector designs were evaluated. One swept-ramp model had 10-deg compression-ramps and the other had 10-deg expansion cavities between flush wall ramps. The scramjet combustor model was instrumented with pressure taps and heat-flux gages. The pressure measurements indicated that both injector configurations were effective in promoting mixing and combustion. Autoignition occurred for the compression-ramp injectors, and the fuel began to burn immediately downstream of the injectors. In tests of the expansion ramps, a pilot was required to ignite the fuel, and the fuel did not burn for a distance of at least two gaps downstream of the injectors. Once initiated, combustion was rapid in this configuration. Heat transfer measurements showed that the heat flux differed greatly both across the width of the combustor and along the length of the combustor.

  6. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  7. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    CERN Document Server

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  8. Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boafo, Emmanuel [Ghana atomic Energy Commission, Accra (Ghana)

    2013-07-01

    The burnup capability in the MCNPX code was utilized to perform fuel depletion analysis of the MNSR LEU core by estimating the amount of fissile material (U-235) consumed as well as the amount of plutonium formed after the reactor core expected life. The decay heat removal rate for the MNSR after reactor shutdown was also investigated due to its significance to reactor safety. The results show that 0.568 % of U-235 was burnt up after 200 days of reactor operation while the amount of plutonium formed was not significant. The study also found that the decay heat decreased exponentially after reactor shutdown confirming that the decay heat will be removed from the system by natural circulation after shut down and hence safety of the reactor is assured.

  9. Development, implementation, and verification of multicycle depletion perturbation theory for reactor burnup analysis

    Energy Technology Data Exchange (ETDEWEB)

    White, J.R.

    1980-08-01

    A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.

  10. 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.

  11. Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis

    Science.gov (United States)

    Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.

    2012-12-01

    The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant α. We have found out, that for the rate of α change, the inequality ěrt˙ {α }/α ěrt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.

  12. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    Science.gov (United States)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  13. Investigation of the fundamental constants stability based on the reactor Oklo burn-up analysis

    CERN Document Server

    Onegin, M S

    2010-01-01

    The burn-up for SC56-1472 sample of the natural Oklo reactor zone 3 was calculated using the modern Monte Carlo codes. We reconstructed the neutron spectrum in the core by means of the isotope ratios: $^{147}$Sm/$^{148}$Sm and $^{176}$Lu/$^{175}$Lu. These ratios unambiguously determine the spectrum index and core temperature. The effective neutron absorption cross section of $^{149}$Sm calculated using this spectrum was compared with experimental one. The disagreement between these two values allows to limit a possible shift of the low laying resonance of $^{149}$Sm even more . Then, these limits were converted to the limits for the change of the fine structure constant $\\alpha$. We found that for the rate of $\\alpha$ change the inequality $|\\delta \\dot{\\alpha}/\\alpha| \\le 5\\cdot 10^{-18}$ is fulfilled, which is of the next higher order than our previous limit.

  14. 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.

  15. Generation of Optical Vortices by Linear Phase Ramps

    Directory of Open Access Journals (Sweden)

    Sunil Vyas

    2012-01-01

    Full Text Available Generation of optical vortices using linear phase ramps is experimentally demonstrated. When two regions of a wavefront have opposite phase gradients then along the line of phase discontinuity vortices can be generated. It is shown that vortices can evolve during propagation even with the unequal magnitude of tilt in the two regions of the wavefront. The number of vortices and their location depend upon the magnitude of tilt. vortex generation is experimentally realized by encoding phase mask on spatial light modulator and their presence is detected interferometrically. Numerical simulation has been performed to calculate the diffracted intensity distribution from the phase mask, and presence of vortices in the diffracted field is detected by computational techniques.

  16. [Ramp lesions : Tips and tricks in diagnostics and therapy].

    Science.gov (United States)

    Seil, R; Hoffmann, A; Scheffler, S; Theisen, D; Mouton, C; Pape, D

    2017-09-14

    There is an increasing biomechanical and anatomical understanding of the different types of meniscal lesions. Lesions of the posterior part of the medial meniscus in the meniscosynovial area have recently received increased attention. They generally occur in association with anterior cruciate ligament (ACL) injuries. They are often missed ("hidden lesions") due to the fact that they cannot be seen by routine anterior arthroscopic inspection. Furthermore, meniscosynovial lesions play a role in anteroposterior knee laxity and, as such, they may be a cause of failure of ACL reconstruction or of postoperative persistent laxity. Little information is available regarding their cause with respect to injury mechanism, natural history, biomechanical implications, healing potential and treatment options. This article presents an overview of the currently available knowledge of these ramp lesions, their possible pathomechanism, classification, biomechanical relevance as well as repair techniques.

  17. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  18. 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.

  19. Current ramp-up with lower hybrid current drive in EAST

    Science.gov (United States)

    Ding, B. J.; Li, M. H.; Fisch, N. J.; Qin, H.; Li, J. G.; Wilson, J. R.; Kong, E. H.; Zhang, L.; Wei, W.; Li, Y. C.; Wang, M.; Xu, H. D.; Gong, X. Z.; Shen, B.; Liu, F. K.; Shan, J. F.

    2012-12-01

    More economical fusion reactors might be enabled through the cyclic operation of lower hybrid current drive. The first stage of cyclic operation would be to ramp up the plasma current with lower hybrid waves alone in low-density plasma. Such a current ramp-up was carried out successfully on the EAST tokamak. The plasma current was ramped up with a time-averaged rate of 18 kA/s with lower hybrid (LH) power. The average conversion efficiency Pel/PLH was about 3%. Over a transient phase, faster ramp-up was obtained. These experiments feature a separate measurement of the L/R time at the time of current ramp up.

  20. Potential Field Cellular Automata Model for Pedestrian Evacuation in a Domain with a Ramp

    Directory of Open Access Journals (Sweden)

    Xiao-Xia Jian

    2014-01-01

    Full Text Available We propose a potential field cellular automata model with a pushing force field to simulate the pedestrian evacuation in a domain with a ramp. We construct a cost potential depending on the ramp angle and introduce a function to evaluate the pushing force, which is related to the cost and the desired direction of pedestrian. With increase of crowd density, there is no empty space for pedestrian moving forward; pedestrian will purposefully push another pedestrian on her or his desired location to arrive the destination quickly. We analyse the relationship between the slope of ramp and the pushing force and investigate the changing of injured situations with the changing of the slope of ramp. When the number of pedestrians and the ramp angle arrive at certain critical points, the Domino effect will be simulated by this proposed model.

  1. Generation of ramp waves using variable areal density flyers

    Science.gov (United States)

    Winter, R. E.; Cotton, M.; Harris, E. J.; Chapman, D. J.; Eakins, D.

    2016-07-01

    Ramp loading using graded density impactors as flyers in gas-gun-driven plate impact experiments can yield new and useful information about the equation of state and the strength properties of the loaded material. Selective Laser Melting, an additive manufacturing technique, was used to manufacture a graded density flyer, termed the "bed-of-nails" (BON). A 2.5-mm-thick × 99.4-mm-diameter solid disc of stainless steel formed a base for an array of tapered spikes of length 5.5 mm and spaced 1 mm apart. The two experiments to test the concept were performed at impact velocities of 900 and 1100 m/s using the 100-mm gas gun at the Institute of Shock Physics at Imperial College London. In each experiment, a BON flyer was impacted onto a copper buffer plate which helped to smooth out perturbations in the wave profile. The ramp delivered to the copper buffer was in turn transmitted to three tantalum targets of thicknesses 3, 5 and 7 mm, which were mounted in contact with the back face of the copper. Heterodyne velocimetry (Het-V) was used to measure the velocity-time history, at the back faces of the tantalum discs. The wave profiles display a smooth increase in velocity over a period of ˜ 2.5 μs, with no indication of a shock jump. The measured profiles have been analysed to generate a stress vs. volume curve for tantalum. The results have been compared with the predictions of the Sandia National Laboratories hydrocode, CTH.

  2. Transport and Burnup Numerical Simulation on the Liquid Blanket Burnup of In-Zinerater%In-Zinerater液态包层输运燃耗数值模拟

    Institute of Scientific and Technical Information of China (English)

    师学明; 杨俊云; 刘成安

    2014-01-01

    Z-Pinch惯性约束聚变是未来一种有竞争力的能源候选方案。Z-Pinch驱动的聚变裂变混合堆可高效地嬗变反应堆乏燃料中分离出的超铀元素。对美国Sandia国家实验室提出的In-Zinerater混合堆概念进行了中子学分析和数值模拟。在三维输运燃耗耦合程序MCORGS中增加了处理在线添加燃料与去除裂变产物的功能,实现了对液态燃料燃耗过程的模拟。增加6Li丰度和燃料初装量保持寿期初反应性不变,可以减缓寿期内反应性下降趋势。逐步增加包层内超铀元素装量,可以控制整个寿期内反应性基本恒定。聚变功率取20 MW,通过反应性控制,5年内包层能量放大倍数在160∼180之间,氚增殖比在1.5∼1.7之间,优于In-Zinerater基准设计方案。%Z-Pinch Inertial confinement fusion is a competitive candidate for future energy solution. A fusion-fission hybrid driven by Z-Pinch can be used to transmute transuranic elements from spent fuels of reactors efficiently. Analysis and numerical simulation of blanket neutronics of In-Zinerater, which is a fusion-fission hybrid concept design in Sandia National Laboratories, is given in this paper. Modification to the three dimension transport and burnup code MCORGS are done, so as to simulate continuous feeding and continuous chemical processing of the liquid fuel. Different combination of initial enrichment of 6Li and fuels loading in the blanket are selected to keep the same reactivity at begin of core. By this way, the decreasing trend of reactivity at life of the core can be lowered. The reactivity can be maintained constant by increasing the fuel loading in the core gradually as the burnup deepens. Given a 20 MW fusion power, by reactivity control, the blanket energy multiplication is around 160∼180 and tritium breed ratio 1.5∼1.7 in 5 years, which is a better result than Sandia’s original design.

  3. Tracking at LHC as a collaborative data challenge use case with RAMP

    CERN Document Server

    Amrouche, Sabrina; Calafiura, Paolo; Farrell, Steven; Gammler, Jochen; Germain, Cécile; Gligorov, Vladimir Vava; Golling, Tobias; Grasland, Hadrien; Gray, Heather; Guyon, Isabelle; Hushchyn, Mikhail; Innocente, Vincenzo; Kégl, Balázs; Neuhaus, Sara; Rousseau, David; Salzburger, Andreas; Ustyuzhanin, Andrei; Vlimant, Jean-Roch; Wessel, Christian; Yilmaz, Yetkin

    2017-01-01

    Charged particle tracking has been a major component of data-processing in high-energy physics experiments such as the Large Hadron Collider (LHC), and is fore- seen to become more and more challenging. There are many ways to perform the tracking task; a collaborative platform, RAMP, has been set up so that developers can create algo- rithms to solve a simplified 2D tracking problems. A small scale competition was held during Connecting The Dots / Intelligent Trackers 2017 (CTDWIT 2017) workshop. De- spite the short time scale, a number of different approaches have been exercised. The scoring algorithm which has been developed, was shown to be robust both during the competition and with stress tests performed after the competition.

  4. Non-isentropic layers in matter behind shock and ramp compression waves

    CERN Document Server

    Khishchenko, Konstantin V

    2014-01-01

    According to the ideal fluid dynamics approach, the temperature and entropy values of a medium undergo a jump increase in the shock front as well as on contact interface between different materials after the shock wave propagation, but remain constant behind the shock front out of the contact interface. In the real condensed matter, the shock fronts and transition regions near the interfaces have finite thicknesses; therefore, the temperature field is disturbed around the interfaces. In this work, such disturbances are numerically analyzed for the problems of formation of the steady shock wave at impact and ramp loading of metals, reflection of the steady shock wave from a free surface, and the shock wave passing through the interface between two different materials. Theoretical analysis and computations show that the non-isentropic layers (the high-entropy ones with the increased temperature and the low-entropy ones with the decreased temperature) arise near the interfaces in the above problems of shock and ...

  5. A power ramped pulsed mode laser piercing technique for improved CO 2 laser profile cutting

    Science.gov (United States)

    Tirumala Rao, B.; Ittoop, M. O.; Kukreja, L. M.

    2009-11-01

    Laser piercing is one of the inevitable requirements of laser profile cutting process and it has a direct bearing on the quality of the laser cut profiles. We have developed a novel power ramped pulsed mode (PRPM) laser piercing technique to produce much finer pierced holes and to achieve a better control on the process parameters compared to the existing methodology based on normal pulsed mode (NPM). Experiments were carried out with both PRPM and NPM laser piercing on 1.5-mm-thick mild steel using an in-house developed high-power transverse flow continuous wave (CW)-CO 2 laser. Significant improvements in the spatter, circularity of the pierced hole and reproducibility were achieved through the PRPM technique. We studied, in detail, the dynamics of processes involved in PRPM laser piercing and compared that with those of the NPM piercing.

  6. OXYGEN UPTAKE KINETICS DURING INCREMENTAL- AND DECREMENTAL-RAMP CYCLE ERGOMETRY

    Directory of Open Access Journals (Sweden)

    Fadıl Özyener

    2011-09-01

    Full Text Available The pulmonary oxygen uptake (VO2 response to incremental-ramp cycle ergometry typically demonstrates lagged-linear first-order kinetics with a slope of ~10-11 ml·min-1·W-1, both above and below the lactate threshold (ӨL, i.e. there is no discernible VO2 slow component (or "excess" VO2 above ӨL. We were interested in determining whether a reverse ramp profile would yield the same response dynamics. Ten healthy males performed a maximum incremental -ramp (15-30 W·min-1, depending on fitness. On another day, the work rate (WR was increased abruptly to the incremental maximum and then decremented at the same rate of 15-30 W.min-1 (step-decremental ramp. Five subjects also performed a sub-maximal ramp-decremental test from 90% of ӨL. VO2 was determined breath-by-breath from continuous monitoring of respired volumes (turbine and gas concentrations (mass spectrometer. The incremental-ramp VO2-WR slope was 10.3 ± 0.7 ml·min-1·W-1, whereas that of the descending limb of the decremental ramp was 14.2 ± 1.1 ml·min-1·W-1 (p < 0.005. The sub-maximal decremental-ramp slope, however, was only 9. 8 ± 0.9 ml·min-1·W-1: not significantly different from that of the incremental-ramp. This suggests that the VO2 response in the supra-ӨL domain of incremental-ramp exercise manifest not actual, but pseudo, first-order kinetics

  7. Development and validation of burnup dependent computational schemes for the analysis of assemblies with advanced lattice codes

    Science.gov (United States)

    Ramamoorthy, Karthikeyan

    The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant

  8. Study on wake structure characteristics of a slotted micro-ramp with large-eddy simulation

    Science.gov (United States)

    Dong, Xiangrui; Chen, Yaohui; Dong, Gang; Liu, Yixin

    2017-06-01

    In this paper, a novel slotted ramp-type micro vortex generator (slotted micro-ramp) for flow separation control is simulated in the supersonic flow of Ma = 1.5, based on large eddy simulation combined with the finite volume method. The wake structure characteristics and control mechanisms of both slotted and standard micro-ramps are presented and discussed. The results show that the wake of standard micro-ramp includes a primary counter-rotating streamwise vortex pair, a train of vortex rings, and secondary vortices. The slotted micro-ramp has more complicated wake structures, which contain a confluent counter-rotating streamwise vortex pair and additional streamwise vortices, with the same rotation generated by slot and the vortex rings enveloping the vortex pair. The additional vortices generated by the slot of the micro-ramp can mix with the primary counter-rotating vortex pair, extend the life time, and strengthen the vortex intensity of primary vortex pair. Moreover, the slot can effectively alleviate, or even eliminate the backflow and decrease the profile drag induced by the standard micro-ramp, therefore improving the efficiency of separation control.

  9. Deficiency of RAMP1 attenuates antigen-induced airway hyperresponsiveness in mice.

    Directory of Open Access Journals (Sweden)

    Manyu Li

    Full Text Available Asthma is a chronic inflammatory disease affecting the lung, characterized by breathing difficulty during an attack following exposure to an environmental trigger. Calcitonin gene-related peptide (CGRP is a neuropeptide that may have a pathological role in asthma. The CGRP receptor is comprised of two components, which include the G-protein coupled receptor, calcitonin receptor-like receptor (CLR, and receptor activity-modifying protein 1 (RAMP1. RAMPs, including RAMP1, mediate ligand specificity in addition to aiding in the localization of receptors to the cell surface. Since there has been some controversy regarding the effect of CGRP on asthma, we sought to determine the effect of CGRP signaling ablation in an animal model of asthma. Using gene-targeting techniques, we generated mice deficient for RAMP1 by excising exon 3. After determining that these mice are viable and overtly normal, we sensitized the animals to ovalbumin prior to assessing airway resistance and inflammation after methacholine challenge. We found that mice lacking RAMP1 had reduced airway resistance and inflammation compared to wildtype animals. Additionally, we found that a 50% reduction of CLR, the G-protein receptor component of the CGRP receptor, also ameliorated airway resistance and inflammation in this model of allergic asthma. Interestingly, the loss of CLR from the smooth muscle cells did not alter the airway resistance, indicating that CGRP does not act directly on the smooth muscle cells to drive airway hyperresponsiveness. Together, these data indicate that signaling through RAMP1 and CLR plays a role in mediating asthma pathology. Since RAMP1 and CLR interact to form a receptor for CGRP, our data indicate that aberrant CGRP signaling, perhaps on lung endothelial and inflammatory cells, contributes to asthma pathophysiology. Finally, since RAMP-receptor interfaces are pharmacologically tractable, it may be possible to develop compounds targeting the RAMP1/CLR

  10. RAMP1 suppresses mucosal injury from dextran sodium sulfate-induced colitis in mice.

    Science.gov (United States)

    Kawashima-Takeda, Noriko; Ito, Yoshiya; Nishizawa, Nobuyuki; Kawashima, Rei; Tanaka, Kiyoshi; Tsujikawa, Kazutake; Watanabe, Masahiko; Majima, Masataka

    2017-04-01

    Calcitonin gene-related peptide (CGRP) is thought to be involved in the modulation of intestinal motility. CGRP receptor is composed of receptor activity-modifying protein (RAMP) 1 combined with calcitonin receptor-like receptor (CRLR) for CGRP. The study investigated the role of CGRP in mice with experimentally induced colitis. The study used dextran sodium sulfate (DSS) to induce colitis in mice. The study compared the severity of colitis in wild-type (WT) mice, mice treated with a CGRP receptor antagonist (CGRP8-37 ), and RAMP1 knockout ((-/-) ) mice. Pathological changes in the mucosa were assessed, and inflammatory cells and cytokine levels were measured. The severity of inflammation in DSS-induced colitis increased markedly in CGRP8-37 -treated mice and RAMP1(-/-) mice compared with WT mice. RAMP1(-/-) mice showed more severe damage compared with CGRP8-37 -treated mice. The number of periodic acid-Schiff-positive cells decreased in CGRP8-37 -treated mice compared with WT mice and was even further decreased in RAMP1(-/-) mice. RAMP1 was expressed by macrophages, mast cells, and T-cells. RAMP1(-/-) mice exhibited excessive accumulation of macrophages and mast cells into the colonic tissue with increased levels of tumor necrosis factor-α and interleukin-1β as compared with WT mice. Infiltration of T-cells into the colonic mucosa, which was associated with the expression of T helper (Th) cytokines including Th1 (interferon gamma) and Th17 (IL-17), was augmented in RAMP1(-/-) mice. The findings of this study suggest that RAMP1 exerted mucosal protection in DSS-induced colitis via attenuation of recruitment of inflammatory cells and of pro-inflammatory cytokines. © 2016 Journal of Gastroenterology and Hepatology Foundation and John Wiley & Sons Australia, Ltd.

  11. Comparator circuits with local ramp buffering for a column-parallel single slope ADC

    Energy Technology Data Exchange (ETDEWEB)

    Milkov, Mihail M.

    2016-04-26

    A comparator circuit suitable for use in a column-parallel single-slope analog-to-digital converter comprises a comparator, an input voltage sampling switch, a sampling capacitor arranged to store a voltage which varies with an input voltage when the sampling switch is closed, and a local ramp buffer arranged to buffer a global voltage ramp applied at an input. The comparator circuit is arranged such that its output toggles when the buffered global voltage ramp exceeds the stored voltage. Both DC- and AC-coupled comparator embodiments are disclosed.

  12. The Joint Control of Variable Speed and On-Ramp Metering for Urban Expressway

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The traffic performance of urban expressway is subject to non-recurring and recurring events, which may cause heavy congestion and vehicles long queuing on ramps. The low performance may bring more traffic delay to the whole network of urban road. This paper presents a new method, the joint control of variable speed control and on-ramp metering, which attempts to improve the level of traffic operations on urban expressway. By analyzing traffic flow on urban expressway, an optimum control strategy of variable speed and on-ramp metering is established in the paper.``

  13. Parametric study of single expansion ramp nozzles at subsonic/transonic speeds

    Science.gov (United States)

    Capone, F. J.; Re, R. J.; Bare, E. A.; Maclean, M. K.

    1987-01-01

    The Langley Research Center has conducted a parametric investigation to determine the aeropropulsive characteristics of single expansion ramp nozzles (SERN). The SERN is a nonaxisymmetric, variable-area, internal/external expansion exhaust nozzle. Internal nozzle parameters that were varied included upper ramp length, ramp chordal angle, lower flap length, flap angle and the axial and vertical locations of nozzle throat. Convergent-divergent and convergent nozzles were included in this investigation which was conducted in the Langley 16-Foot Transonic Tunnel at Mach numbers from 0.6 to 1.2 and at nozzle pressure ratios up to 12.0.

  14. Evaluation technology for burnup and generated amount of plutonium by measurement of xenon isotopic ratio in dissolver off-gas at reprocessing facility (Joint research)

    OpenAIRE

    岡野 正紀; 久野 剛彦; 高橋 一朗; 白水 秀知; Charlton, W. S.; Wells, C. A.; Hemberger, P. H.; 山田 敬二; 酒井 敏雄

    2006-01-01

    The amount of Pu in the spent fuel was evaluated from Xe isotopic ratio in off-gas in reprocessing facility, is related to burnup. Six batches of dissolver off-gas at spent fuel dissolution process were sampled from the main stack in Tokai Reprocessing Plant during BWR fuel reprocessing campaign. Xenon isotopic ratio was determined with GC/MS. Burnup and generated amount of Pu were evaluated with Noble Gas Environmental Monitoring Application code (NOVA), developed by Los Alamos National Labo...

  15. Burnup determination of a fuel element concerning different cooling times; Seguimiento del quemado de un elemento combustible, para diferentes tiempos de enfriamento

    Energy Technology Data Exchange (ETDEWEB)

    Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O. [Comision Chilena de Energia Nuclear, Santiago (Chile). Dept. de Aplicaciones Nucleares. Unidad de Reactores; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)

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

    Directory of Open Access Journals (Sweden)

    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.

  17. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    Science.gov (United States)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  18. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U 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.

  19. 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

  20. Current applications of actinide-only burn-up credit within the Cogema group and R and D programme to take fission products into account

    Energy Technology Data Exchange (ETDEWEB)

    Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.

  1. 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.

  2. Micro-Ramps for External Compression Low-Boom Inlets

    Science.gov (United States)

    Rybalko, Michael; Loth, Eric; Chima, Rodrick V.; Hirt, Stefanie M.; DeBonis, James R.

    2010-01-01

    The application of vortex generators for flow control in an external compression, axisymmetric, low-boom concept inlet was investigated using RANS simulations with three-dimensional (3-D), structured, chimera (overset) grids and the WIND-US code. The low-boom inlet design is based on previous scale model 1- by 1-ft wind tunnel tests and features a zero-angle cowl and relaxed isentropic compression centerbody spike, resulting in defocused oblique shocks and a weak terminating normal shock. Validation of the methodology was first performed for micro-ramps in supersonic flow on a flat plate with and without oblique shocks. For the inlet configuration, simulations with several types of vortex generators were conducted for positions both upstream and downstream of the terminating normal shock. The performance parameters included incompressible axisymmetric shape factor, separation area, inlet pressure recovery, and massflow ratio. The design of experiments (DOE) methodology was used to select device size and location, analyze the resulting data, and determine the optimal choice of device geometry. The optimum upstream configuration was found to substantially reduce the post-shock separation area but did not significantly impact recovery at the aerodynamic interface plane (AIP). Downstream device placement allowed for fuller boundary layer velocity profiles and reduced distortion. This resulted in an improved pressure recovery and massflow ratio at the AIP compared to the baseline solid-wall configuration.

  3. SOUTH RAMP 3.01.X AREA GROUND SUPPORT ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    S. Bonabian

    1999-07-12

    The purpose of this analysis is to evaluate the stability and determine ground support requirements for the 3.01.X areas in the Exploratory Studies Facility (ESF) South Ramp. The 3.01.X area refers to the ESF tunnel portions that were constructed under Section 3.01.X of the ESF General Construction Specification (Reference 8.4). Four 3.01.X areas in the ESF Main Loop are covered in this analysis that extend from Station 60+15.28 to 60+49.22, 62+04.82 to 62+32.77, 75+21.02 to 75+28.38, and 76+63.08 to 77+41.23. The scope of the analysis is (1) to document the as-built configuration including existing voids and installed ground support, (2) to evaluate the existing ground conditions, (3) to determine applicable design loads, (4) to evaluate the stability and determine a ground support system, and (5) to analyze the recommended system.

  4. LHC Report: intensity ramp-up – familiar demons

    CERN Multimedia

    Mike Lamont for the LHC team

    2015-01-01

    The first 2015 scrubbing run ended on Friday, 3 July and successfully delivered a well-scrubbed machine ready for operation with a 50 ns beam. This opened the way for the first phase of the so-called beam intensity ramp-up. The last couple of weeks have seen the number of bunches increase from 3 to 476 per beam via periods of 50, 144 and 300 bunches per beam.   The graph plots the rate of LHC beam dumps due to single-event effects (SEE) versus beam luminosity. It is an indication of the importance of tackling this issue.   To verify the full and proper functioning of all systems, operators need at least 3 fills and 20 hours of stable beams without significant problems. After 20 hours, an extensive checklist is signed off by the system experts before the next step up in the number of bunches. The systems involved include magnet protection, radio-frequency, beam instrumentation, collimation, operations, feedback, beam dump and injection. Increasing the total beam intensity poses a numb...

  5. 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.

  6. Characterization of the non-uniqueness of used nuclear fuel burnup signatures through a Mesh-Adaptive Direct Search

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E., E-mail: sskutnik@utk.edu; Davis, David R.

    2016-05-01

    The use of passive gamma and neutron signatures from fission indicators is a common means of estimating used fuel burnup, enrichment, and cooling time. However, while characteristic fission product signatures such as {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, and others are generally reliable estimators for used fuel burnup within the context where the assembly initial enrichment and the discharge time are known, in the absence of initial enrichment and/or cooling time information (such as when applying NDA measurements in a safeguards/verification context), these fission product indicators no longer yield a unique solution for assembly enrichment, burnup, and cooling time after discharge. Through the use of a new Mesh-Adaptive Direct Search (MADS) algorithm, it is possible to directly probe the shape of this “degeneracy space” characteristic of individual nuclides (and combinations thereof), both as a function of constrained parameters (such as the assembly irradiation history) and unconstrained parameters (e.g., the cooling time before measurement and the measurement precision for particular indicator nuclides). In doing so, this affords the identification of potential means of narrowing the uncertainty space of potential assembly enrichment, burnup, and cooling time combinations, thereby bounding estimates of assembly plutonium content. In particular, combinations of gamma-emitting nuclides with distinct half-lives (e.g., {sup 134}Cs with {sup 137}Cs and {sup 154}Eu) in conjunction with gross neutron counting (via {sup 244}Cm) are able to reasonably constrain the degeneracy space of possible solutions to a space small enough to perform useful discrimination and verification of fuel assemblies based on their irradiation history.

  7. 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)

  8. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

  9. Optimized Swinging Door Algorithm for Wind Power Ramp Event Detection: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Mingjian; Zhang, Jie; Florita, Anthony R.; Hodge, Bri-Mathias; Ke, Deping; Sun, Yuanzhang

    2015-08-06

    Significant wind power ramp events (WPREs) are those that influence the integration of wind power, and they are a concern to the continued reliable operation of the power grid. As wind power penetration has increased in recent years, so has the importance of wind power ramps. In this paper, an optimized swinging door algorithm (SDA) is developed to improve ramp detection performance. Wind power time series data are segmented by the original SDA, and then all significant ramps are detected and merged through a dynamic programming algorithm. An application of the optimized SDA is provided to ascertain the optimal parameter of the original SDA. Measured wind power data from the Electric Reliability Council of Texas (ERCOT) are used to evaluate the proposed optimized SDA.

  10. Stability and scour development of bed material on crossbar block ramps

    Institute of Scientific and Technical Information of China (English)

    Mario OerteLa; DanieLB Bung

    2015-01-01

    Block ramps are ecologically oriented drop structures with adequate energy dissipation and partially moderate flow velocities. A special case is given with crossbar block ramps, where the upstream and downstream level difference is reduced by a series of basins. To prevent the total structure from failing, the stability of single boulders within the crossbars and the bed material in between must be guaranteed. The present paper addresses the stability of bed material and scour development for various flow regimes. Any bed material erosion may affect the stability of the crossbar boulders, which in turn can result in major damages of the ramp. Therefore new design approaches are developed to choose an appropriate bed material size and to avoid failures of crossbar block ramp structures.

  11. MEASURING CHROMATICITY ALONG THE RAMP USING THE PLL TUNE METER IN RHIC.

    Energy Technology Data Exchange (ETDEWEB)

    TEPIKIAN,S.; AHRENS,L.; CAMERON,P.; SCHULTHEISS,C.

    2002-06-02

    Beam stability up the ramp requires the appropriate sign and magnitude of the chromaticity. We developed a way to measure the chromaticity using the PLL (Phase Locked Loop) tune-meter. Since, the accuracy of the PLL tune-meter with properly adjusted loop gain is better than {approx} 0.0001 in tune units, the radial loop needs only be changed by a small amount of 0.2mm at a 1Hz rate. Thus, we can achieve fast chromaticity measurements in 1 sec. Except during the very beginning of the ramp where there are snapback effects and the gamma changes very rapidly, we can have good chromaticcity measurements along the ramp. This leads to the possibility of correcting the chromaticity during the ramp using a feedback system.

  12. Modelling ramp-up curves to reflect learning: improving capacity planning in secondary pharmaceutical production

    DEFF Research Database (Denmark)

    Hansen, Klaus Reinholdt Nyhuus; Grunow, Martin

    2015-01-01

    are overestimated. We develop a new method, which captures ramp-up as a function of the cumulative production volume to better reflect the experience gained while producing the new product. The use of the more accurate and computationally effective approach is demonstrated for the case of secondary pharmaceutical...... production. Due to its regulatory framework, this industry cannot fully exploit available capacities during ramp-up. We develop a capacity planning model for a new pharmaceutical drug, which determines the number and location of new production lines and the build-up of inventory such that product...... availability at market launch is ensured. Our MILP model is applied to a real industry case study using three empirically observed ramp-up curves to demonstrate its value as decision support tool. We demonstrate the superiority of our volume-dependent method over the traditional time-dependent ramp...

  13. First beam test of a combined ramp and squeeze at LHC

    CERN Document Server

    Wenninger, Jorg; Coello De Portugal - Martinez Vazquez, Jaime Maria; Gorzawski, Arkadiusz; Redaelli, Stefano; Schaumann, Michaela; Solfaroli Camillocci, Matteo; CERN. Geneva. ATS Department

    2015-01-01

    With increasing maturity of LHC operation it is possible to envisage more complex beam manipulations. At the same time operational efficiency receives increasing attention. So far ramping the beams to their target energy and squeezing the beams to smaller or higher beta are decoupled at the LHC. (De-)squeezing is always performed at the target energy, currently 6.5 TeV. Studies to combine the ramp and squeeze processes have been made for the LHC since 2011, but so far no experimental test with beam had ever performed. This note describes the first machine experiment with beam aiming at validating the combination of ramp and squeeze, the so-called combined ramp and squeeze (CRS).

  14. Tamping Ramping: Algorithmic, Implementational, and Computational Explanations of Phasic Dopamine Signals in the Accumbens.

    Directory of Open Access Journals (Sweden)

    Kevin Lloyd

    2015-12-01

    Full Text Available Substantial evidence suggests that the phasic activity of dopamine neurons represents reinforcement learning's temporal difference prediction error. However, recent reports of ramp-like increases in dopamine concentration in the striatum when animals are about to act, or are about to reach rewards, appear to pose a challenge to established thinking. This is because the implied activity is persistently predictable by preceding stimuli, and so cannot arise as this sort of prediction error. Here, we explore three possible accounts of such ramping signals: (a the resolution of uncertainty about the timing of action; (b the direct influence of dopamine over mechanisms associated with making choices; and (c a new model of discounted vigour. Collectively, these suggest that dopamine ramps may be explained, with only minor disturbance, by standard theoretical ideas, though urgent questions remain regarding their proximal cause. We suggest experimental approaches to disentangling which of the proposed mechanisms are responsible for dopamine ramps.

  15. Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

    Science.gov (United States)

    Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

    2014-12-01

    Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

  16. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan

    1999-09-01

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  17. A state of the Art report on Manufacturing technology of high burn-up fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan

    1999-09-01

    In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. (author)

  18. MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP Obrigheim

    Energy Technology Data Exchange (ETDEWEB)

    Cao Yan, E-mail: ycao@anl.go [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Broeders, Cornelis H.M. [Forschungszentrum Karlsruhe, Institute for Neutron Physics and Reactor Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2010-10-15

    This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

  19. Development and verification of fuel burn-up calculation model in a reduced reactor geometry

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, Tagor Malem [Center for Reactor Technology and Nuclear Safety (PTKRN), National Nuclear Energy Agency (BATAN), Kawasan PUSPIPTEK Gd. No. 80, Serpong, Tangerang 15310 (Indonesia)], E-mail: tagorms@batan.go.id; Liem, Peng Hong [Research Laboratory for Nuclear Reactor (RLNR), Tokyo Institute of Technology (Tokyo Tech), O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2008-02-15

    A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning the applicability of the interpolation functions, it can be concluded that the optimal coefficients of the interpolation functions are not very sensitive to the core configuration and core or fuel composition in RSG GAS (MPR-30) reactor. Consequently, once the optimal interpolation function and its coefficients are derived then they can be used for 2-D routine operational in-core fuel management without repeating the expensive 3-D neutron diffusion calculations. At the selected fuel elements (at H-9 and G-6 core grid positions), the discrepancy of the FECFs (fuel element channel power peaking factors) between the 2-D and 3-D models are within the range of 3.637 x 10{sup -4}, 3.241 x 10{sup -4} and 7.556 x 10{sup -4} for the oxide, silicide cores with 250 g {sup 235}U/FE and the silicide core with 300 g {sup 235}U/FE, respectively.

  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. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    CERN Document Server

    Chiesa, Davide; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2015-01-01

    A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...

  2. Propagation of nuclear data uncertainties for ELECTRA burn-up calculations

    CERN Document Server

    ostrand, H; Duan, J; Gustavsson, C; Koning, A; Pomp, S; Rochman, D; Osterlund, M

    2013-01-01

    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 Pu-239 transport data to uncertainties in the fuel inventory of ELECTRA during the reactor life using the Total Monte Carlo approach (TMC). Within the TENDL project the nuclear models input parameters were randomized within their uncertainties and 740 Pu-239 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 in the ...

  3. The Line Operations Safety Audit Program: Transitioning From Flight Operations to Maintenance and Ramp Operations

    Science.gov (United States)

    2011-09-01

    A Line Operations Safety Audit ( LOSA ) is a voluntary safety program that collects safety data during normal airline operations and was originally...maintenance and ramp operations. This report provides a review of the use of LOSA , discusses LOSA’s essential operating characteristics, lessons learned on...the flight deck, and describes the extension of LOSA to maintenance and ramp operations. The research team developed tools for airlines and

  4. Properties of Phase Transition of Traffic Flow on Urban Expressway Systems with Ramps and Accessory Roads

    Institute of Scientific and Technical Information of China (English)

    梅超群; 刘业进

    2011-01-01

    In this paper, we develop a cellular automaton model to describe the phase transition of traffic flow on urban expressway systems with on-off-ramps and accessory roads. The lane changing rules are given in detailed, the numerical results show that the main road and the accessory road both produce phase transitions. These phase transitions will omen be influenced by the number of lanes, lane changing, the ramp flow, the input flow rate, and the geometry structure.

  5. Probabilistic Description of Traffic Breakdowns Caused by On-ramp Flow

    OpenAIRE

    Kuhne, Reinhart; Lubashevsky, Ihor; Mahnke, Reinhard; Kaupush, Jevgenijs

    2004-01-01

    The characteristic features of traffic breakdown near on-ramp are analyzed. To describe this phenomenon the probabilistic description regarding the jam emergence as the formation of a large car cluster on highway inside the synchronized traffic is constructed. In these terms the breakdown occurs through the formation of a certain critical nucleus in the metastable vehicle flow, which is located near the on-ramp. The strong cooperative car interaction in the synchronized traffic enables us to ...

  6. Analyzing the Impact of Solar Power on Multi-Hourly Thermal Generator Ramping

    Energy Technology Data Exchange (ETDEWEB)

    Rosenkranz, Joshua-Benedict; Brancucci Martinez-Anido, Carlo; Hodge, Bri-Mathias

    2016-04-08

    Solar power generation, unlike conventional forms of electricity generation, has higher variability and uncertainty in its output because solar plant output is strongly impacted by weather. As the penetration rate of solar capacity increases, grid operators are increasingly concerned about accommodating the increased variability and uncertainty that solar power provides. This paper illustrates the impacts of increasing solar power penetration on the ramping of conventional electricity generators by simulating the operation of the Independent System Operator -- New England power system. A production cost model was used to simulate the power system under five different scenarios, one without solar power and four with increasing solar power penetrations up to 18%, in terms of annual energy. The impact of solar power is analyzed on six different temporal intervals, including hourly and multi-hourly (2- to 6-hour) ramping. The results show how the integration of solar power increases the 1- to 6-hour ramping events of the net load (electric load minus solar power). The study also analyzes the impact of solar power on the distribution of multi-hourly ramping events of fossil-fueled generators and shows increasing 1- to 6-hour ramping events for all different generators. Generators with higher ramp rates such as gas and oil turbine and internal combustion engine generators increased their ramping events by 200% to 280%. For other generator types--including gas combined-cycle generators, coal steam turbine generators, and gas and oil steam turbine generators--more and higher ramping events occurred as well for higher solar power penetration levels.

  7. Downstream Effect of Ramping Neuronal Activity through Synapses with Short-Term Plasticity.

    Science.gov (United States)

    Wei, Wei; Wang, Xiao-Jing

    2016-04-01

    Ramping neuronal activity refers to spiking activity with a rate that increases quasi-linearly over time. It has been observed in multiple cortical areas and is correlated with evidence accumulation processes or timing. In this work, we investigated the downstream effect of ramping neuronal activity through synapses that display short-term facilitation (STF) or depression (STD). We obtained an analytical result for a synapse driven by deterministic linear ramping input that exhibits pure STF or STD and numerically investigated the general case when a synapse displays both STF and STD. We show that the analytical deterministic solution gives an accurate description of the averaging synaptic activation of many inputs converging onto a postsynaptic neuron, even when fluctuations in the ramping input are strong. Activation of a synapse with STF shows an initial cubical increase with time, followed by a linear ramping similar to a synapse without STF. Activation of a synapse with STD grows in time to a maximum before falling and reaching a plateau, and this steady state is independent of the slope of the ramping input. For a synapse displaying both STF and STD, an increase in the depression time constant from a value much smaller than the facilitation time constant τ(F) to a value much larger than τ(F) leads to a transition from facilitation dominance to depression dominance. Therefore, our work provides insights into the impact of ramping neuronal activity on downstream neurons through synapses that display short-term plasticity. In a perceptual decision-making process, ramping activity has been observed in the parietal and prefrontal cortices, with a slope that decreases with task difficulty. Our work predicts that neurons downstream from such a decision circuit could instead display a firing plateau independent of the task difficulty, provided that the synaptic connection is endowed with short-term depression.

  8. Particle Confinement Improvement with Current Ramp in the HL-IM Tokamak

    Institute of Scientific and Technical Information of China (English)

    YAN Long-Wen; WANG En-Yao; QIAN Jun; CHEN Liao-Yuan; LIU Yong

    2003-01-01

    Particle confinement with current ramp is investigated in the HL-IM tokamak. The ratio of particle confinement time to energy confinement time is used to determine confinement improvement. It increases a factor of 3 after the current rises twofold. An optimization density range to improve particle confinement is (1.5-3.5) XlO19 m~3. Particle confinement improvement during current ramp-up is beneficial to the startup of fusion reactors.

  9. A 4 MA, 500 ns pulsed power generator CQ-4 for characterization of material behaviors under ramp wave loading

    Science.gov (United States)

    Wang, Guiji; Luo, Binqiang; Zhang, Xuping; Zhao, Jianheng; Sun, Chengwei; Tan, Fuli; Chong, Tao; Mo, Jianjun; Wu, Gang; Tao, Yanhui

    2013-01-01

    A pulsed power generator CQ-4 was developed to characterize dynamic behaviors of materials under ramp wave loading, and to launch high velocity flyer plates for shock compression and hypervelocity impact experiments of materials and structures at Institute of Fluid Physics, China Academy of Engineering Physics. CQ-4 is composed of twenty capacitor and primary discharge switch modules with total capacitance of 32μF and rated charging voltage of 100 kV, and the storage energy is transmitted by two top and bottom parallel aluminum plates insulated by twelve layers of polyester film with total thickness of 1.2 mm. Between capacitor bank and chamber, there are 72 peaking capacitors with total capacitance of 7.2 μF and rated voltage of 120 kV in parallel, which are connected with the capacitor bank in parallel. Before the load, there is a group of seven secondary self-breaking down switches connected with the total circuit in series. The peaking capacitors and secondary switches are used to shape the discharging current waveforms. For short-circuit, the peak current of discharging can be up to 3 ˜ 4 MA and rise time varies from 470 ns to 600 ns when the charging voltages of the generator are from 75 kV to 85 kV. With CQ-4 generator, some quasi-isentropic compression experiments under ramp wave loadings are done to demonstrate the ability of CQ-4 generator. And some experiments of launching high velocity flyer plates are also done on CQ-4. The experimental results show that ramp wave loading pressure of several tens of GPa on copper and aluminum samples can be realized and the velocity of aluminum flyer plate with size of 10 mm × 6 mm × 0.35 mm can be accelerated to about 11 km/s and the velocity of aluminum flyer plate with size of 10 mm × 6 mm × 0.6 mm can be up to about 9 km/s, which show that CQ-4 is a good and versatile tool to realize ramp wave loading and shock compression for shock physics.

  10. A 4 MA, 500 ns pulsed power generator CQ-4 for characterization of material behaviors under ramp wave loading.

    Science.gov (United States)

    Wang, Guiji; Luo, Binqiang; Zhang, Xuping; Zhao, Jianheng; Sun, Chengwei; Tan, Fuli; Chong, Tao; Mo, Jianjun; Wu, Gang; Tao, Yanhui

    2013-01-01

    A pulsed power generator CQ-4 was developed to characterize dynamic behaviors of materials under ramp wave loading, and to launch high velocity flyer plates for shock compression and hypervelocity impact experiments of materials and structures at Institute of Fluid Physics, China Academy of Engineering Physics. CQ-4 is composed of twenty capacitor and primary discharge switch modules with total capacitance of 32 μF and rated charging voltage of 100 kV, and the storage energy is transmitted by two top and bottom parallel aluminum plates insulated by twelve layers of polyester film with total thickness of 1.2 mm. Between capacitor bank and chamber, there are 72 peaking capacitors with total capacitance of 7.2 μF and rated voltage of 120 kV in parallel, which are connected with the capacitor bank in parallel. Before the load, there is a group of seven secondary self-breaking down switches connected with the total circuit in series. The peaking capacitors and secondary switches are used to shape the discharging current waveforms. For short-circuit, the peak current of discharging can be up to 3 ~ 4 MA and rise time varies from 470 ns to 600 ns when the charging voltages of the generator are from 75 kV to 85 kV. With CQ-4 generator, some quasi-isentropic compression experiments under ramp wave loadings are done to demonstrate the ability of CQ-4 generator. And some experiments of launching high velocity flyer plates are also done on CQ-4. The experimental results show that ramp wave loading pressure of several tens of GPa on copper and aluminum samples can be realized and the velocity of aluminum flyer plate with size of 10 mm × 6 mm × 0.35 mm can be accelerated to about 11 km/s and the velocity of aluminum flyer plate with size of 10 mm × 6 mm × 0.6 mm can be up to about 9 km/s, which show that CQ-4 is a good and versatile tool to realize ramp wave loading and shock compression for shock physics.

  11. Ramp Forecasting Performance from Improved Short-Term Wind Power Forecasting: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, J.; Florita, A.; Hodge, B. M.; Freedman, J.

    2014-05-01

    The variable and uncertain nature of wind generation presents a new concern to power system operators. One of the biggest concerns associated with integrating a large amount of wind power into the grid is the ability to handle large ramps in wind power output. Large ramps can significantly influence system economics and reliability, on which power system operators place primary emphasis. The Wind Forecasting Improvement Project (WFIP) was performed to improve wind power forecasts and determine the value of these improvements to grid operators. This paper evaluates the performance of improved short-term wind power ramp forecasting. The study is performed for the Electric Reliability Council of Texas (ERCOT) by comparing the experimental WFIP forecast to the current short-term wind power forecast (STWPF). Four types of significant wind power ramps are employed in the study; these are based on the power change magnitude, direction, and duration. The swinging door algorithm is adopted to extract ramp events from actual and forecasted wind power time series. The results show that the experimental short-term wind power forecasts improve the accuracy of the wind power ramp forecasting, especially during the summer.

  12. Deformation in thrust-ramp anticlines and duplexes: implications for geometry and porosity

    Energy Technology Data Exchange (ETDEWEB)

    Groshong, R.H. Jr.; Usdansky, S.I.

    1986-05-01

    A computerized kinematic model of thrust-ramp anticline geometry allows workers to predict the zones of greatest deformation in ramp anticlines and fault duplexes. The model assumes a constant cross-section area, symmetrical fold hinges, and slip in the hanging wall parallel to the ramp and forelimb. Assuming that the collapse of original porosity or the generation of secondary fracture porosity is proportional to deformation, the model can be used to predict porosity changes. Deformation in a single ramp anticline is greatest in the forelimb and backlimb, and may be absent in the crest. A duplex structure results from comparatively closely spaced thrusts that have a common upper detachment horizon. Relatively wide spacing between the duplex faults yields a bumpy roofed duplex as in the central Appalachians. Forelimbs may be deformed twice and should show greater porosity modification. Relatively close spacing between ramp-and-flat thrusts can produce a listric-fault, snakehead anticline geometry because younger faults deform the preexisting thrust slices. The resulting geometry is here called a snakehead duplex and appears to be fairly common, as in the Jumpingpound field in the Canadian Rockies. Each thrust slice within the duplex is deformed six times or more, providing the maximum opportunity for deformation-related porosity changes. Maximum fracture porosity should occur in thrusts having listric-fan or snakehead duplex geometry. Structures involving duplexes generally should be better than isolated ramp anticlines.

  13. Cellular automata model simulating traffic car accidents in the on-ramp system

    Science.gov (United States)

    Echab, H.; Lakouari, N.; Ez-Zahraouy, H.; Benyoussef, A.

    2015-01-01

    In this paper, using Nagel-Schreckenberg model we study the on-ramp system under the expanded open boundary condition. The phase diagram of the two-lane on-ramp system is computed. It is found that the expanded left boundary insertion strategy enhances the flow in the on-ramp lane. Furthermore, we have studied the probability of the occurrence of car accidents. We distinguish two types of car accidents: the accident at the on-ramp site (Prc) and the rear-end accident in the main road (Pac). It is shown that car accidents at the on-ramp site are more likely to occur when traffic is free on road A. However, the rear-end accidents begin to occur above a critical injecting rate αc1. The influence of the on-ramp length (LB) and position (xC0) on the car accidents probabilities is studied. We found that large LB or xC0 causes an important decrease of the probability Prc. However, only large xC0 provokes an increase of the probability Pac. The effect of the stochastic randomization is also computed.

  14. Engineering and design of holding yards, loading ramps and handling facilities for land and sea transport of livestock

    Directory of Open Access Journals (Sweden)

    Temple Grandin

    2008-03-01

    Full Text Available Facilities designed for intensively raised animals trained to lead are not appropriate for handling extensively raised animals unaccustomed to close contact with people. The author provides information on facility design for both intensively and extensively raised livestock. Non-slip flooring in handling facilities is essential for all livestock. Cleats must be spaced on loading ramps for trucks or ships so that the hooves of the animals fit easily between them. Cleats spaced too far apart cause slipping and falling. In developing countries, building stationary ramps for vehicles of differing heights using concrete, wood or steel is recommended. Highly mechanised systems, such as hydraulic tailgate lifts, are not recommended in developing countries due to maintenance difficulties. The holding capacity for livestock shipping and receiving terminals should be designed to hold the largest number of animals handled on the busiest days. To maintain high standards of animal welfare, it is important to train employees to handle animals using methods to reduce stress and to conduct weekly audits of handling using an objective, numerical scoring system to maintain high welfare standards.

  15. Late Miocene breccia of Menorca (Balearic Islands) a basis for the interpretation of a Neogene ramp deposit

    Science.gov (United States)

    Obrador, A.; Pomar, L.; Taberner, C.

    1992-08-01

    Neogene (Tortonian) ramp facies associations crop out in the southern sector of Menorca (Balearic island). These are made mainly by sigmoidal and oblique clinoform units comprising rhodoliths, bryozoans, molluscs and foraminifera. These units are interpreted as outer-ramp deposits. A breccia deposit infilling an erosional surface is found at the top of the carbonate ramp sequence. The breccia components (mainly rhodoliths and oolite clasts) may represent erosion of the underlying ramp deposits and inner-ramp counterparts, which do not crop out on the island. The study of components in the breccia deposits confirms the indigenous character of the outer-ramp facies associations, which suggests that the ramp was steepened distally. The breccia deposits correspond laterally to a discontinuity surface locally showing karstic features. Transgressive sediments (including a phosphatic crust) are found above the discontinuity surface. All together these features, and the dolomitisation of the uppermost ramp sediments and breccia deposits, suggest that the breccia originated from erosion of the ramp after a major relative sea-level fall. The breccia and discontinuity surface separate the ramp sequence from an overlying prograding sequence. A correlation of this sequence boundary to other areas in the western Mediterranean is proposed.

  16. CD uniformity optimization at volume ramp up stage for new product introduction

    Science.gov (United States)

    Kim, Jin-Soo; Ma, Won-Kwang; Kim, Young-Sik; Kim, Myoung-Soo; Kwon, Won-Taik; Park, Sung-Ki; Nikolsky, Peter; Otter, Marian; Marun, Maryana Escalante; Anunciado, Roy; Sun, Kyu-Tae; Storms, Greet; van West, Ewould

    2014-04-01

    In this paper we describe the joint development and optimization of the critical dimension uniformity (CDU) at an advanced 300 mm ArFi semiconductor facility of SK Hynix in the high volume device. As the ITRS CDU specification shrinks, semiconductor companies still need to maintain high wafer yield and high performance (hence market value) even during the introduction phase of a new product. This cannot be achieved without continuous improvement of the on-product CDU as one of the main drivers for yield improvement. ASML Imaging Optimizer is one of the most efficient tools to reach this goal. This paper presents experimental results of post-etch CDU improvement by ASML imaging optimizer for immature photolithography and etch processes on critical features of 20nm node. We will show that CDU improvement potential and measured CDU strongly depend on CD fingerprint stability through wafers, lots and time. However, significant CDU optimization can still be achieved, even for variable CD fingerprints. In this paper we will review point-to-point correlation of CD fingerprints as one of the main indicators for CDU improvement potential. We will demonstrate the value of this indicator by comparing CD correlation between wafers used for Imaging Optimizer dose recipe development, predicted and measured CDU for wafers and lots exposed with various delays ranging from a few days to a month. This approach to CDU optimization helps to achieve higher yield earlier in the new product introduction cycle, enables faster technology ramps and thereby improves product time to market.

  17. Coarse cross-bedded grainstones in a mid- to outer carbonate ramp, Bartonian of the Urbasa-Andia plateau (W Pyrenees, N Spain)

    Science.gov (United States)

    Baceta, José I.; Pomar, Luis; Mateu-Vicens, Guillem

    2017-04-01

    Most marine grainstones in carbonate ramps and platforms are commonly interpreted to form in high-energy, shallow-water settings where wave energy dissipates by friction on the sea floor. The locus of energy dissipation varies with platform type. On rimmed shelves, skeletal-oolitic sands mainly accumulate near the wave-agitated shelf margin as a rim, which restricts wave action and a low-energy lagoon may form landwards. On ramps and open platforms, by contrast, grainstones commonly accumulate in the shallower zone near or attached to the shoreline, grading basinward into muddier carbonate successions. Within this conceptual scheme, most carbonate ramp subdivisions have been established according to the facies and sedimentary structures associated to the bathymetry-related hydraulic regime, in which the bases of the surface storm-waves and the fairweather waves are the boundary layers. Since seagrasses encroached the oceans by the late Cretaceous, baffling the surface wave energy, and burrowing activity increased significantly, most Cenozoic ramp successions lack the bathymetry-related sedimentary structures and the record of wave and storm activity is commonly lost. This has induced ramp subdivision to become progressively based in light penetration, as inferred from the light dependence of the carbonate producers, particularly for the Cenozoic. This new scenario has permitted to recognize grainstone units detached from shoreline and shoals and produced at depths near the limit of light penetration, or even below, in basinal settings. Here we document a 90-100-m thick Eocene example of crossbedded skeletal grainstones composed by echinoderm-, bryozoan-, red-algal fragments and orthophragminid larger benthic foraminifers. This facies belt occurs at 20-km from the paleo-coastline, downdip of Nummulites-Discocyclina facies, and passes basinward into finely comminuted skeletal debris and marls with planktonic foraminifers of the outer ramp. The skeletal composition of

  18. Source Term Analysis for Reactor Coolant System with Consideration of Fuel Burnup

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yu Jong; Ahn, Joon Gi; Hwang, Hae Ryong [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    The radiation source terms in reactor coolant system (RCS) of pressurized water reactor (PWR) are basic design information for ALARA design such as radiation protection and shielding. Usually engineering companies own self-developed computer codes to estimate the source terms in RCS. DAMSAM and FIPCO are the codes developed by engineering companies. KEPCO E and C has developed computer code, RadSTAR, for use in the Radiation Source Term Analysis for Reactor coolant system during normal operation. The characteristics of RadSTAR are as follows. (1) RadSTAR uses fuel inventory data calculated by ORIGEN, such as ORIGEN2 or ORIGEN-S to consider effects of the fuel burnup. (2) RadSTAR estimates fission products by using finite differential method and analytic method to minimize numerical error. (3) RadSTAR enhances flexibility by adding the function to build the nuclide data library (production pathway library) for user-defined nuclides from ORIGEN data library. (4) RadSTAR consists of two modules. RadSTAR-BL is to build the nuclide data library. RadSTAR-ST is to perform numerical analysis on source terms. This paper includes descriptions on the numerical model, the buildup of nuclide data library, and the sensitivity analysis and verification of RadSTAR. KEPCO E and C developed RadSTAR to calculate source terms in RCS during normal operation. Sensitivity analysis and accuracy verification showed that RadSTAR keeps stability at Δt of 0.1 day and gives more accurate results in comparison with DAMSAM. After development, RadSTAR will replace DAMSAM. The areas, necessary to further development of RadSTAR, are addition of source term calculations for activation products and for shutdown operation.

  19. Sequence development of a latest Devonian-Tournaisian distally-steepened mixed carbonate-siliciclastic ramp, Canning Basin, Australia

    Science.gov (United States)

    Seyedmehdi, Zahra; George, Annette D.; Tucker, Maurice E.

    2016-03-01

    The sequence development and evolution of latest Devonian-earliest Carboniferous Fairfield Group in the Canning Basin have been established through integration of detailed sedimentological analysis of core, petrophysical data, existing biostratigraphic data and new seismic interpretations. The Fairfield Group on the Lennard Shelf was deposited on a mixed carbonate-siliciclastic distally-steepened ramp with a broad inner ramp, narrow mid ramp and steepened outer ramp. The majority of facies associations (FA1-FA8) were formed in intertidal-shallow subtidal conditions in proximal to distal inner ramp including siliciclastic tidal flats (FA1), carbonate intertidal flats (FA2), tidal flats and channels (FA3), lagoons (FA4-FA5), and shallow subtidal (FA6), backshoal (FA7) and fore-shoal areas (FA8). Bioclastic muddy sandstone (FA9) and bioclastic mudstone (FA10) are the dominant mid-ramp facies. Recognition of turbiditic facies of middle to lower slope of the outer ramp (FA11-FA13) led to the identification of a distally-steepened ramp. Antecedent topography exerted a significant control on platform morphology and the development of the widespread inner-ramp facies on the Lennard Shelf. A sequence-stratigraphic analysis reveals that the Fairfield Group ramp deposits consists of four third-order sequences (S1-S4) that were largely deposited during sea-level highstands (HST) characterized by progradational trends and dominant shallow subtidal inner-ramp facies associations. Transgressive systems tracts (TST) are well developed in S1 and S3 and have a retrogradational facies pattern with dominant deep subtidal mid-outer ramp facies associations. Lowstand systems tracts, characterized by lowstand wedges and turbiditic facies, are identified in the lower parts of S2 and S3. Coarse and fine-grained siliciclastic facies are mixed with carbonate facies as a result of coeval deposition on the inner and mid ramp, and reciprocal deposition on the outer ramp. A temporal variation in

  20. Ramp-wave compression experiment with direct laser illumination on Shen Guang III prototype Laser facility

    Science.gov (United States)

    Wang, Feng; Xu, Tao; Optical Team Collaboration

    2016-10-01

    Ramp-wave compression (RWC) experiment to balance the high compression pressure generation in aluminum and x-ray blanking effect in transparent window was demonstrated on Shen Guang-III prototype laser facility. A new target concept was proposed to develop a laser-driven shocks-RWC technique for studying material behavior under dynamic, high pressure conditions. As the ``little shocks'' in our experiment cannot be avoided, the effort to diminish the shock under a special level has been demonstrated with Al/Au/Al/LiF target. The highest pressure is about 500GPa after using the multilayer target design Al/Au/Al/LiF and about 1013W/cm2 laser pulse incident on the planer Al target, instantaneously affecting ablation layer located 500 μm away. As the x-ray generated by Al layer had been prevented by the Au layer, the width abrupt onset of strong absorption of an optical probe beam (λ = 532 nm) in LiF window may be the limitation for this kind if RWC experiment during the experiment time scale for 30 μm thick step. With the design laser shape and target structure of Al/Au/Al/LiF, 500GPa may be the highest pressure after balance the preheat effect and ablation efficiency for laser direct-drive experiment.

  1. Geotechnical characterization of the North Ramp of the Exploratory Studies Facility: Yucca Mountain Site Characterization Project. Volume 2, NRG corehole data appendices

    Energy Technology Data Exchange (ETDEWEB)

    Brechtel, C.E.; Lin, Ming; Martin, E. [Agapito Associates, Inc., Grand Junction, CO (United States); Kessel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

    1995-05-01

    This report presents the results of the geological and geotechnical characterization of the Miocene volcanic tuff rocks of the Timber Mountain and Paintbrush groups that the tunnel boring machine will encounter during excavations of the Exploratory Studies Facility (ESF) North Ramp. The information in this report was developed to support the design of the ESF North Ramp. The ESF is being constructed by the DOE as part of the Yucca Mountain Project site characterization activities. The purpose of these activities is to evaluate the potential to locate the national high-level nuclear waste repository on land within and adjacent to the Nevada Test Site (NTS), Nye County, Nevada. This report was prepared as part of the Soil and Rock Properties Studies in accordance with the 8.3.1.14.2 Study Plan to Provide Soil and Rock Properties. This is volume 2 which contains NRG Corehole Data for each of the NRG Holes.

  2. Numerical Simulation on Ramp Initiation and Propagation in a Fold-and-thrust Belt and Accretionary Wedge

    Science.gov (United States)

    Hu, C.; Liu, X.; Shi, Y.

    2015-12-01

    Fold-and-thrust belts and accretionary wedge develop along compressive plate boundaries, both in hinterland and foreland. Under the long-term compressive tectonic loading, a series ramps will initiate and propagate along the wedge. How do the ramps initiate? What are the timing and spacing intervals between the ramps? How many patterns are there for the ramp propagation? These questions are basic for the study of ramp initiation and propagation. Many scholars used three different methods, critical coulomb wedge theory, analogue sandbox models, and numerical simulation to research the initiation and propagation of the ramps, respectively. In this paper, we set up a 2-D elastic-plastic finite element model, with a frictional contact plane, to simulate the initiation and propagation of the ramps. In this model, the material in upper wedge is homogenous, but considering the effects of gravity and long-term tectonic loading. The model is very simple but simulated results are very interesting. The simulated results indicate that the cohesion of upper wedge and dip angle of detachment plane have strong effects on the initiation and propagation of ramps. There are three different patterns of ramp initiation and propagation for different values of the cohesion. The results are different from those by previous analogue sandbox models, and numerical simulation, in which there is usually only one pattern for the ramp initiation and propagation. The results are consistent with geological survey for the ramp formation in an accretionary wedge. This study will provide more knowledge of mechanism of the ramp initiation and propagation in Tibetan Plateau and central Taiwan.

  3. 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.

  4. Capacity Estimation for On-Ramp Merging Section of Urban Expressway Based on Time Headway Loss

    Directory of Open Access Journals (Sweden)

    Xing-jian Xue

    2014-01-01

    Full Text Available This paper proposes a model for estimating capacity of on-ramp merging section of urban expressway based on dynamics and gap acceptance theory, considering lane-changing processes and time headway loss. Survey data were collected from on-ramp merging sections of shanghai urban expressway system and showed that capacity drop of on-ramp merging section is caused by drivers’ lane-changing which may lead to unsteady speed of vehicles and so prolonged time headway compared to the minimum time headway corresponding to the maximum capacity. Three parameters (optimal time headway, time headway loss, and interference quantity of lane-changing are given and a methodology by accumulating time headway loss due to lane-changing is developed to estimate the capacity drop. Results’ comparisons between real data and microsimulation of on-ramp merging sections and sensitivity analysis show that the proposed model can produce reliable and accurate results. This study also reveals that ramp flow and the difference between the optimal speed and the lane-changing speed of fleet have a great impact on capacity drop. This study is beneficial to evaluate congestion levels, to understand complex traffic phenomena, and so to find efficient solutions.

  5. Solar Power Ramp Events Detection Using an Optimized Swinging Door Algorithm: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Mingjian; Zhang, Jie; Florita, Anthony; Hodge, Bri-Mathias; Ke, Deping; Sun, Yuanzhang

    2015-08-07

    Solar power ramp events (SPREs) are those that significantly influence the integration of solar power on non-clear days and threaten the reliable and economic operation of power systems. Accurately extracting solar power ramps becomes more important with increasing levels of solar power penetrations in power systems. In this paper, we develop an optimized swinging door algorithm (OpSDA) to detection. First, the swinging door algorithm (SDA) is utilized to segregate measured solar power generation into consecutive segments in a piecewise linear fashion. Then we use a dynamic programming approach to combine adjacent segments into significant ramps when the decision thresholds are met. In addition, the expected SPREs occurring in clear-sky solar power conditions are removed. Measured solar power data from Tucson Electric Power is used to assess the performance of the proposed methodology. OpSDA is compared to two other ramp detection methods: the SDA and the L1-Ramp Detect with Sliding Window (L1-SW) method. The statistical results show the validity and effectiveness of the proposed method. OpSDA can significantly improve the performance of the SDA, and it can perform as well as or better than L1-SW with substantially less computation time.

  6. Probabilistic Swinging Door Algorithm as Applied to Photovoltaic Power Ramping Event Detection

    Energy Technology Data Exchange (ETDEWEB)

    Florita, Anthony; Zhang, Jie; Brancucci Martinez-Anido, Carlo; Hodge, Bri-Mathias; Cui, Mingjian

    2015-10-02

    Photovoltaic (PV) power generation experiences power ramping events due to cloud interference. Depending on the extent of PV aggregation and local grid features, such power variability can be constructive or destructive to measures of uncertainty regarding renewable power generation; however, it directly influences contingency planning, production costs, and the overall reliable operation of power systems. For enhanced power system flexibility, and to help mitigate the negative impacts of power ramping, it is desirable to analyze events in a probabilistic fashion so degrees of beliefs concerning system states and forecastability are better captured and uncertainty is explicitly quantified. A probabilistic swinging door algorithm is developed and presented in this paper. It is then applied to a solar data set of PV power generation. The probabilistic swinging door algorithm builds on results from the original swinging door algorithm, first used for data compression in trend logging, and it is described by two uncertain parameters: (i) e, the threshold sensitivity to a given ramp, and (ii) s, the residual of the piecewise linear ramps. These two parameters determine the distribution of ramps and capture the uncertainty in PV power generation.

  7. The constant work rate critical power protocol overestimates ramp incremental exercise performance.

    Science.gov (United States)

    Black, Matthew I; Jones, Andrew M; Kelly, James A; Bailey, Stephen J; Vanhatalo, Anni

    2016-12-01

    The parameters of the power-duration relationship (i.e., the critical power, CP, and the curvature constant, W') may theoretically predict maximal performance capability for exercise above the CP. The CP and W' are associated with the parameters of oxygen uptake ([Formula: see text]O2) kinetics, which can be altered by manipulation of the work-rate forcing function. We tested the hypothesis that the CP and W' derived from constant work-rate (CWR) prediction trials would overestimate ramp incremental exercise performance. Thirty subjects (males, n = 28; females, n = 2) performed a ramp incremental test, and 3-5 CWR prediction trials for the determination of the CP and W'. Multiple ramp incremental tests and corresponding CP and W' estimates were available for some subjects such that in total 51 ramp test performances were predicted. The ramp incremental test performance (729 ± 113 s) was overestimated by the CP and W' estimates derived from the best (751 ± 114 s, P incremental performance suggests that the CP and W' derived from different work-rate forcing functions, thus resulting in different [Formula: see text]O2 kinetics, cannot be used interchangeably. The present findings highlight a potential source of error in performance prediction that is of importance to both researchers and applied practitioners.

  8. Elution parameters in constant-pressure, single-ramp temperature-programmed gas chromatography.

    Science.gov (United States)

    Blumberg, L M; Klee, M S

    2001-05-18

    The dependence of the degree of interaction of a solute with the stationary phase at the time of its elution from the column in temperature-programmed GC is best described by interaction level of the solute. The latter represents the fraction of a solute residing in the stationary phase relative to the total amount of the solute. A simple approach to the evaluation of interaction levels of eluting solutes in a single-ramp temperature program is proposed. In a single-ramp temperature program having no preceding temperature plateau, all solutes that elute at temperatures that are about 60 degrees C higher than the initial temperature of the heating ramp elute with nearly the same interaction levels that can be found as exp(-r), where r is dimensionless heating rate. A specially designed temperature plateau preceding the ramp causes all solutes eluting during the entire time of the ramp to elute with nearly the same interaction levels equal to exp(-r). A transformation of the interaction level of a solute into its retention factor or mobility factor (a fraction of a solute in a mobile phase in relation to the total amount of the solute) and vice versa is also described.

  9. A New Macro Model for Traffic Flow on a Highway with Ramps and Numerical Tests

    Institute of Scientific and Technical Information of China (English)

    TANG Tie-Qiao; HUANG Hai-Jun; S.C.Wong; GAO Zi-You; ZHANG Ying

    2009-01-01

    In this paper, we present a new macro model for traffic flow on a highway with ramps based on the existing models. We use the new model to study the effects of on-off-ramp on the main road traffic during the morning rush period and the evening rush period. Numerical tests show that, during the two rush periods, these effects are often different and related to the status of the main road traffic. If the main road traffic flow is uniform, then ramps always produce stop-and-go traffic when the main road density is between two critical values, and ramps have little effect on the main road traffic when the main road density is less than the smaller critical value or greater than the larger critical value. If a small perturbation appears on the main road, ramp may lead to stop-and-go traffic, or relieve or even eliminate the stop-and-go traffic, under different circumstances. These results are consistent with real traffic, which shows that the new model is reasonable.

  10. Ramp-preserving denoising for conductivity image reconstruction in magnetic resonance electrical impedance tomography.

    Science.gov (United States)

    Lee, Chang-Ock; Jeon, Kiwan; Ahn, Seonmin; Kim, Hyung Joong; Woo, Eung Je

    2011-07-01

    In magnetic resonance electrical impedance tomography, among several conductivity image reconstruction algorithms, the harmonic B(z) algorithm has been successfully applied to B(z) data from phantoms and animals. The algorithm is, however, sensitive to measurement noise in B(z) data. Especially, in in vivo animal and human experiments where injection current amplitudes are limited within a few milliampere at most, measured B(z) data tend to have a low SNR. In addition, magnetic resonance (MR) signal void in outer layers of bones and gas-filled organs, for example, produces salt-pepper noise in the MR phase and, consequently, B(z) images. The B(z) images typically present areas of sloped transitions, which can be assimilated to ramps. Conductivity contrasts change ramp slopes in B(z) images and it is critical to preserve positions of those ramps to correctly recover edges in conductivity images. In this paper, we propose a ramp-preserving denoising method utilizing a structure tensor. Using an eigenvalue analysis, we identified local regions of salt-pepper noise. Outside the identified local regions, we applied an anisotropic smoothing to reduce noise while preserving their ramp structures. Inside the local regions of salt-pepper noise, we used an isotropic smoothing. After validating the proposed denoising method through numerical simulations, we applied it to in vivo animal imaging experiments. Both numerical simulation and experimental results show significant improvements in the quality of reconstructed conductivity images. © 2011 IEEE

  11. Climatological attribution of wind power ramp events in East Japan and their probabilistic forecast based on multi-model ensembles downscaled by analog ensemble using self-organizing maps

    Science.gov (United States)

    Ohba, Masamichi; Nohara, Daisuke; Kadokura, Shinji

    2016-04-01

    Severe storms or other extreme weather events can interrupt the spin of wind turbines in large scale that cause unexpected "wind ramp events". In this study, we present an application of self-organizing maps (SOMs) for climatological attribution of the wind ramp events and their probabilistic prediction. The SOM is an automatic data-mining clustering technique, which allows us to summarize a high-dimensional data space in terms of a set of reference vectors. The SOM is applied to analyze and connect the relationship between atmospheric patterns over Japan and wind power generation. SOM is employed on sea level pressure derived from the JRA55 reanalysis over the target area (Tohoku region in Japan), whereby a two-dimensional lattice of weather patterns (WPs) classified during the 1977-2013 period is obtained. To compare with the atmospheric data, the long-term wind power generation is reconstructed by using a high-resolution surface observation network AMeDAS (Automated Meteorological Data Acquisition System) in Japan. Our analysis extracts seven typical WPs, which are linked to frequent occurrences of wind ramp events. Probabilistic forecasts to wind power generation and ramps are conducted by using the obtained SOM. The probability are derived from the multiple SOM lattices based on the matching of output from TIGGE multi-model global forecast to the WPs on the lattices. Since this method effectively takes care of the empirical uncertainties from the historical data, wind power generation and ramp is probabilistically forecasted from the forecasts of global models. The predictability skill of the forecasts for the wind power generation and ramp events show the relatively good skill score under the downscaling technique. It is expected that the results of this study provides better guidance to the user community and contribute to future development of system operation model for the transmission grid operator.

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

    Science.gov (United States)

    Nemtsev, G.; Amosov, V.; Meshchaninov, S.; Popovichev, S.; Rodionov, R.

    2016-11-01

    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.

  13. Optimization of ramp area aircraft push back time windows in the presence of uncertainty

    Science.gov (United States)

    Coupe, William Jeremy

    It is well known that airport surface traffic congestion at major airports is responsible for increased taxi-out times, fuel burn and excess emissions and there is potential to mitigate these negative consequences through optimizing airport surface traffic operations. Due to a highly congested voice communication channel between pilots and air traffic controllers and a data communication channel that is used only for limited functions, one of the most viable near-term strategies for improvement of the surface traffic is issuing a push back advisory to each departing aircraft. This dissertation focuses on the optimization of a push back time window for each departing aircraft. The optimization takes into account both spatial and temporal uncertainties of ramp area aircraft trajectories. The uncertainties are described by a stochastic kinematic model of aircraft trajectories, which is used to infer distributions of combinations of push back times that lead to conflict among trajectories from different gates. The model is validated and the distributions are included in the push back time window optimization. Under the assumption of a fixed taxiway spot schedule, the computed push back time windows can be integrated with a higher level taxiway scheduler to optimize the flow of traffic from the gate to the departure runway queue. To enable real-time decision making the computational time of the push back time window optimization is critical and is analyzed throughout.

  14. DNA exit ramps are revealed in the binding landscapes obtained from simulations in helical coordinates.

    Directory of Open Access Journals (Sweden)

    Ignacia Echeverria

    2015-02-01

    Full Text Available DNA molecules are highly charged semi-flexible polymers that are involved in a wide variety of dynamical processes such as transcription and replication. Characterizing the binding landscapes around DNA molecules is essential to understanding the energetics and kinetics of various biological processes. We present a curvilinear coordinate system that fully takes into account the helical symmetry of a DNA segment. The latter naturally allows to characterize the spatial organization and motions of ligands tracking the minor or major grooves, in a motion reminiscent of sliding. Using this approach, we performed umbrella sampling (US molecular dynamics (MD simulations to calculate the three-dimensional potentials of mean force (3D-PMFs for a Na+ cation and for methyl guanidinium, an arginine analog. The computed PMFs show that, even for small ligands, the free energy landscapes are complex. In general, energy barriers of up to ~5 kcal/mol were measured for removing the ligands from the minor groove, and of ~1.5 kcal/mol for sliding along the minor groove. We shed light on the way the minor groove geometry, defined mainly by the DNA sequence, shapes the binding landscape around DNA, providing heterogeneous environments for recognition by various ligands. For example, we identified the presence of dissociation points or "exit ramps" that naturally would terminate sliding. We discuss how our findings have important implications for understanding how proteins and ligands associate and slide along DNA.

  15. DNA exit ramps are revealed in the binding landscapes obtained from simulations in helical coordinates.

    Science.gov (United States)

    Echeverria, Ignacia; Papoian, Garegin A

    2015-02-01

    DNA molecules are highly charged semi-flexible polymers that are involved in a wide variety of dynamical processes such as transcription and replication. Characterizing the binding landscapes around DNA molecules is essential to understanding the energetics and kinetics of various biological processes. We present a curvilinear coordinate system that fully takes into account the helical symmetry of a DNA segment. The latter naturally allows to characterize the spatial organization and motions of ligands tracking the minor or major grooves, in a motion reminiscent of sliding. Using this approach, we performed umbrella sampling (US) molecular dynamics (MD) simulations to calculate the three-dimensional potentials of mean force (3D-PMFs) for a Na+ cation and for methyl guanidinium, an arginine analog. The computed PMFs show that, even for small ligands, the free energy landscapes are complex. In general, energy barriers of up to ~5 kcal/mol were measured for removing the ligands from the minor groove, and of ~1.5 kcal/mol for sliding along the minor groove. We shed light on the way the minor groove geometry, defined mainly by the DNA sequence, shapes the binding landscape around DNA, providing heterogeneous environments for recognition by various ligands. For example, we identified the presence of dissociation points or "exit ramps" that naturally would terminate sliding. We discuss how our findings have important implications for understanding how proteins and ligands associate and slide along DNA.

  16. Observation of Solid-Solid Phase Transitions in Ramp-Compressed Aluminum

    Science.gov (United States)

    Polsin, D. N.; Boehly, T. R.; Delettrez, J. A.; Gregor, M. C.; McCoy, C. A.; Henderson, B.; Fratanduono, D. E.; Smith, R.; Kraus, R.; Eggert, J. H.; Collins, R.; Coppari, F.; Celliers, P. M.

    2016-10-01

    We present results of experiments using x-ray diffraction to study the crystalline structure of solid aluminum compressed up to 500 GPa. Aluminum is of interest because it is frequently used as a standard material in high-pressure compression experiments. At ambient pressure and temperature, Al is a face-centered cubic close-packed crystal and has been observed to transform to hexagonal close-packed (hcp) when compressed to 200GPa in a diamond anvil cell. It is predicted to transform from hcp to body-centered cubic when compressed to 315GPa. Laser-driven ramp waves will be used to compress Al to various constant-pressure states. The goal is to investigate the Al phase diagram along its isentrope, i.e., at temperatures 1000K and pressures ranging from 200 to 500 GPa. X-ray diffraction will be used to measure the crystalline structure of the compressed Al and observe the transformations that occur at various pressures. This material is based upon work supported by the Department of Energy National Nuclear Security Administration under Award Number DE-NA0001944.

  17. Efficient and Accurate Calculation of Burnup Problems with Short-Lived Nuclides by a Krylov Subspace Method with the Newton Divided Difference

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Hee; Cho, Nam Zin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Nowadays lattice physics codes tend to utilize a detailed burnup chain including short-lived nuclides in order to perform more accurate burnup calculations. But, since production codes, for example, ORIGEN2, take account of nuclides which have relatively long half-life, it is inappropriate for such detailed burnup chain calculation. To enhance that drawback, many matrix exponential calculation methods have been developed. Recently, a Krylov subspace method with the PADE approximation was used. In this paper, a Krylov subspace method based on spectral decomposition property of the matrix function theory with the Newton divided difference (NDD) is introduced. It is tested with a sample problem and compared with simple Taylor expansion method

  18. Iterative ramp sharpening for structure signature-preserving simplification of images

    Energy Technology Data Exchange (ETDEWEB)

    Grazzini, Jacopo A [Los Alamos National Laboratory; Soille, Pierre [EC-JRC

    2010-01-01

    In this paper, we present a simple and heuristic ramp sharpening algorithm that achieves local contrast enhancement of vector-valued images. The proposed algorithm performs a local comparison of intensity values as well as gradient strength and directional information derived from the gradient structure tensor so that the sharpening is applied only for pixels found on the ramps around true edges. This way, the contrast between objects and regions separated by a ramp is enhanced correspondingly, avoiding ringing artefacts. It is found that applying this technique in an iterative manner on blurred imagery produces sharpening preserving both structure and signature of the image. The final approach reaches a good compromise between complexity and effectiveness for image simplification, enhancing in an efficient manner the image details and maintaining the overall image appearance.

  19. A historic breakthrough : Kidd Mine sets new mark for underground ramp development

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, R. [Xstrata Copper, Timmins, ON (Canada)

    2010-08-15

    This article discussed a new ventilation system installed as part of an underground ramp development at Xstrata Copper's Kidd Mine. The mine now has the deepest continual ramp from surface in the world. The down ramp is currently exhausted through raises placed on various levels and a permanent 6-metre-diameter raise exhaust system. Ventilation tests are now being conducted to optimize air flow until the breakthrough of a new exhaust raise in 2010. A set of cold stopes and a mechanical refrigeration plant are being used to cool the air. Heat will become easier to manage as the project completes changes to the ventilation set-up. Additional stopes and fresh air raises are scheduled for operation in the first quarter of 2011. The 3-year project has included contributions from several different construction companies. 2 figs.

  20. Non-Cooperative Regulation Coordination Based on Game Theory for Wind Farm Clusters during Ramping Events

    DEFF Research Database (Denmark)

    Qi, Yongzhi; Liu, Yutian; Wu, Qiuwei

    2017-01-01

    of wind farm clusters (WFCs) in order to track scheduled wind power of the WFC during ramping events. In the proposed strategy, a non‐cooperative game is formulated and wind farms compete to provide regulation to the WFC during ramping events. A regulation revenue function is proposed to evaluate......With increasing penetration of wind power in power systems, it is important to track scheduled wind power output as much as possible during ramping events to ensure security of the system. In this paper, a non‐cooperative coordination strategy based on the game theory is proposed for the regulation...... the competition process of wind farms to provide regulation to the WFC which includes revenue of effective regulation (ER), power support regulation and punishment regulation. The multi‐time‐interval Nash equilibrium condition is derived for the regulation competition process of wind farms. By setting parameters...

  1. Tune and Chromaticity Control During Snapback and Ramp in 2015 LHC Operation

    CERN Document Server

    Schaumann, Michaela; Lamont, Mike; Solfaroli Camillocci, Matteo; Todesco, Ezio; Wenninger, Jorg

    2016-01-01

    Because of current redistribution on the superconducting cables, the harmonic components of the magnetic fields of the superconducting magnets in the Large Hadron Collider (LHC) show decay during the low field injection plateau. This results in tune and chromaticity variations for the beams. In the first few seconds of the ramp the original hysteresis state of the magnetic field is restored - the field snaps back. These fast dynamic field changes lead to strong tune and chromaticity excursions that, if not properly controlled, induce beam losses and potentially trigger a beam dump. A feed-forward system applies predicted corrections during the injection plateau and to the first part of the ramp to avoid violent changes of beam conditions. This paper discusses the snapback of tune and chromaticity as observed in 2015, as well as the control of beam parameters during the ramp. It also evaluates the quality of the applied feed-forward corrections and their reproducibility.

  2. Effects of ramp vibrational states on flexural intrinsic vibrations in Besocke-style scanners

    Institute of Scientific and Technical Information of China (English)

    Zhang Hui; Jiang Guo-Zhu; Liu Zhao-Qun; Zhang Shu-Yi; Fan Li

    2013-01-01

    For both the vibrating and steady supporting surfaces of a scanning disk in a Besocke-style piezoelectric scanner,a theoretical model is given by considering the nonlinear lateral friction at the micro-contact interface between the positioning legs and the supporting surface.Numerical simulations demonstrate that unexpected flexural vibrations can arise from a vibrating ramp,and their frequencies are lower than the eigenfrequencies of the scanner in the linearly elastic regime.The vibrations essentially depend on 1) the vibrational states of the supporting ramp and the steel ball tips on the three piezoelectric positioning legs,and 2) the tribological characteristics of the contacts between the tips and the ramp.The results give an insight into the intrinsic vibrations of the scanners,and are applicable in designing and optimizing piezoelectric scanning systems.

  3. Validation of a Ramp Running Protocol for Determination of the True VO2max in Mice

    Science.gov (United States)

    Ayachi, Mohamed; Niel, Romain; Momken, Iman; Billat, Véronique L.; Mille-Hamard, Laurence

    2016-01-01

    In the field of comparative physiology, it remains to be established whether the concept of VO2max is valid in the mouse and, if so, how this value can be accurately determined. In humans, VO2max is generally considered to correspond to the plateau observed when VO2 no longer rises with an increase in workload. In contrast, the concept of VO2peak tends to be used in murine studies. The objectives of the present study were to determine whether (i) a continuous ramp protocol yielded a higher VO2peak than a stepwise, incremental protocol, and (ii) the VO2peak measured in the ramp protocol corresponded to VO2max. The three protocols (based on intensity-controlled treadmill running until exhaustion with eight female FVB/N mice) were performed in random order: (a) an incremental protocol that begins at 10 m.min−1 speed and increases by 3 m.min−1 every 3 min. (b) a ramp protocol with slow acceleration (3 m.min−2), and (c) a ramp protocol with fast acceleration (12 m.min−2). Each protocol was performed with two slopes (0 and 25°). Hence, each mouse performed six exercise tests. We found that the value of VO2peak was protocol-dependent (p VO2max plateau was associated with the fulfillment of two secondary criteria (a blood lactate concentration >8 mmol.l−1 and a respiratory exchange ratio >1). The total duration of the 3 m.min−2 0° ramp protocol was shorter than that of the incremental protocol. Taken as a whole, our results suggest that VO2max in the mouse is best determined by applying a ramp exercise protocol with slow acceleration and no treadmill slope. PMID:27621709

  4. Theory of Feshbach molecule formation in a dilute gas during a magnetic field ramp

    DEFF Research Database (Denmark)

    Williams, J. E.; Nygaard, Nicolai; Clark, C. W.

    2006-01-01

    Starting with coupled atom-molecule Boltzmann equations, we develop a simplified model to understand molecule formation observed in recent experiments. Our theory predicts several key features: (1) the effective adiabatic rate constant is proportional to density; (2) in an adiabatic ramp, the dep......Starting with coupled atom-molecule Boltzmann equations, we develop a simplified model to understand molecule formation observed in recent experiments. Our theory predicts several key features: (1) the effective adiabatic rate constant is proportional to density; (2) in an adiabatic ramp...

  5. State-space approach to vibration of gold nano-beam induced by ramp type heating

    Institute of Scientific and Technical Information of China (English)

    Hamdy M Youssef; Khaled A Elsibai

    2010-01-01

    In the nanoscale beam, two effects become domineering. One is the non-Fourier effect in heat conduction and the other is the coupling effect between temperature and strain rate. In the present study, a generalized solution for the generalized thermoelastic vibration of gold nano-beam resonator induced by ramp type heating is developed. The solution takes into account the above two effects. State-space and Laplace transform methods are used to determine the lateral vibration, the temperature, the displacement, the stress and the strain energy of the beam. The effects of the relaxation time and the ramping time parameters have been studied.

  6. Quantum wave packet dynamics with trajectories: reflections on a downhill ramp potential

    Science.gov (United States)

    Lopreore, Courtney L.; Wyatt, Robert E.

    2000-07-01

    The quantum trajectory method (QTM) for wave packet dynamics involves solving discretized hydrodynamic equations-of-motion in the Lagrangian picture (C. Lopreore, R.E. Wyatt, Phys. Rev. Lett. 82 (1999) 5190). In this Letter, results are presented which illustrate the dynamics of an initial Gaussian wave packet on a downhill ramp potential. Plots are shown for the time evolving probability density, as well as phase space plots and force diagrams. The mechanism, deduced from these plots, surprisingly shows some of the transmitted fluid elements of the wave packet making a U-turn before they head downhill on the ramp potential.

  7. Effect of Air Velocity on Thermal Comfort under Thermal Environment Ramp Changing

    Institute of Scientific and Technical Information of China (English)

    嵇赟喆; 涂光备; 孙琳

    2004-01-01

    Set points of the indoor air temperature and relative humidity in short-term staying location were studied. In this condition, the thermal reaction of human body varied with the ramp changes of the environmental thermal parameters.The change rules of about 60 subjects'thermal reaction to the ramp change of environment were surveyed, and the effect of air movement on the thermal reaction during transient condition was considered by using a questionnaire. With the experimental results and research findings under stable condition, a way to set environmental parameters of short-time staying location was recommended.

  8. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  9. 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.

  10. Experimental characterization and constitutive modeling of the mechanical behavior of molybdenum under electromagnetically applied compression-shear ramp loading

    Science.gov (United States)

    Alexander, C. S.; Ding, J. L.; Asay, J. R.

    2016-03-01

    Magnetically applied pressure-shear (MAPS) is a new experimental technique that provides a platform for direct measurement of material strength at extreme pressures. The technique employs an imposed quasi-static magnetic field and a pulsed power generator that produces an intense current on a planar driver panel, which in turn generates high amplitude magnetically induced longitudinal compression and transverse shear waves into a planar sample mounted on the drive panel. In order to apply sufficiently high shear traction to the test sample, a high strength material must be used for the drive panel. Molybdenum is a potential driver material for the MAPS experiment because of its high yield strength and sufficient electrical conductivity. To properly interpret the results and gain useful information from the experiments, it is critical to have a good understanding and a predictive capability of the mechanical response of the driver. In this work, the inelastic behavior of molybdenum under uniaxial compression and biaxial compression-shear ramp loading conditions is experimentally characterized. It is observed that an imposed uniaxial magnetic field ramped to approximately 10 T through a period of approximately 2500 μs and held near the peak for about 250 μs before being tested appears to anneal the molybdenum panel. In order to provide a physical basis for model development, a general theoretical framework that incorporates electromagnetic loading and the coupling between the imposed field and the inelasticity of molybdenum was developed. Based on this framework, a multi-axial continuum model for molybdenum under electromagnetic loading is presented. The model reasonably captures all of the material characteristics displayed by the experimental data obtained from various experimental configurations. In addition, data generated from shear loading provide invaluable information not only for validating but also for guiding the development of the material model for

  11. Non-timber forest products: ramps in the Waynesville, NC watershed

    Science.gov (United States)

    Kristina Connor; Jim Chamberlain III; Hilliard Gibbs Jr.; Matt Winn

    2015-01-01

    The potential of forest farming was noted as far back as 1929, but the recognition of its importance dates back only 20 to 30 years. The U.S. market for harvested foods and medicinal plants from forests now exceeds $4 billion annually. Ramps (Allium tricoccum Aiton), or wild leeks, grow in patches in the rich moist forests of the eastern United...

  12. Measuring modulated luminescence using non-modulated stimulation: Ramping the sample period

    DEFF Research Database (Denmark)

    Poolton, N.R.J.; Bøtter-Jensen, L.; Andersen, C.E.;

    2003-01-01

    . Directly analogous results to LM-OSL can, however, be achieved with non-modulated excitation sources, by ramping the sample period (RSP) of luminescence detection. RSP-OSL has the distinct advantage over LM-OSL in that, since the excitation remains at full power, data accumulation times (that can...

  13. AN OPTIMAL REPLENISHMENT POLICY FOR DETERIORATING ITEMS WITH RAMP TYPE DEMAND UNDER PERMISSIBLE DELAY IN PAYMENTS

    Directory of Open Access Journals (Sweden)

    Dr. Sanjay Jain

    2010-10-01

    Full Text Available The aim of this paper is to develop an optimal replenishment policy for inventory models of deteriorating items with ramp type demand under permissible delay in payments. Deterioration of items begins on their arrival in stock.  An example is also presented to illustrate the application of developed model.

  14. Phase-ramp reduction in interseismic interferograms from pixel-offsets

    KAUST Repository

    Wang, Teng

    2014-05-01

    Interferometric synthetic aperture radar (InSAR) is increasingly used to measure interseismic deformation. Inaccurate satellite-orbit information, expressed as phase ramps across interseismic interferograms, is believed to be one of the main sources of error in such measurements. However, many interferograms exhibit higher phase gradients than expected from the reported orbital accuracy, suggesting that there are other error sources. Here, we show that interferogram phase ramps are in part caused by uncorrected satellite timing-parameter errors. We propose a two-step approach to reduce the phase ramps using pixel-offsets estimated between SAR amplitude images. The first step involves using a digital elevation model (DEM) to estimate absolute timing-parameter errors for the reference image of the SAR dataset and the second step updates the timing parameters of the master image for each interferogram. We demonstrate a clear ramp reduction on interseismic interferograms covering the North Anatolian Fault in eastern Turkey. The resulting interferograms show clear signs of interseismic deformation even before stacking. © 2014 IEEE.

  15. Combined Ramp and Squeeze to 6.5 TeV in the LHC

    CERN Document Server

    Solfaroli Camillocci, Matteo; Tomás, Rogelio; Wenninger, Jorg

    2016-01-01

    The cycle of the LHC is composed of an energy ramp followed by a betatron squeeze, needed to reduce the beta- star value in the interaction points. Since Run 1, studies have been carried out to investigate the feasibility of combining the two operations, thus considerably reducing the duration of the operational cycle. In Run 2, the LHC is operating at the energy of 6.5 TeV that requires a much longer cycle than that of Run 1. Therefore, the performance gains from a Combined Ramp and Squeeze (CRS) is more interesting. Merging the energy ramp and the betatron squeeze could result in a gain of several minutes for each LHC cycle. With increasing maturity of LHC operation, it is now possible to envisage more complex beam manipulations; this paper describes the first machine experiment with beam, aiming at validating the combination of ramp and squeeze, which was performed in 2015, during a machine development phase. The operation experience with the LHC run at 2.51 TeV, when CRS down to 4 meters was deployed and ...

  16. Modeling Human Dynamics in Combined Ramp-Following and Disturbance-Rejection Tasks

    NARCIS (Netherlands)

    Pool, D.M.; Van Paassen, M.M.; Mulder, M.

    2010-01-01

    This paper investigates the modeling of humanmanual control behavior for pursuit tracking tasks in which target forcing functions consisting of multiple ramp-like changes in target attitude are used. Due to the use of a pursuit display and the predictability of such forcing function signals, it can

  17. The three-dimensional flow organization past a micro-ramp in a supersonic boundary layer

    NARCIS (Netherlands)

    Sun, Z.; Schrijer, F.F.J.; Scarano, F.; Van Oudheusden, B.W.

    2012-01-01

    The three-dimensional instantaneous flow organization in the near wake of a micro-ramp interacting with a Mach 2.0 supersonic turbulent boundary layer is studied using tomographic particle image velocimetry. The mean flow reveals a wake with approximately circular cross section dominated by a pair o

  18. Modeling ramp compression experiments using large-scale molecular dynamics simulation.

    Energy Technology Data Exchange (ETDEWEB)

    Mattsson, Thomas Kjell Rene; Desjarlais, Michael Paul; Grest, Gary Stephen; Templeton, Jeremy Alan; Thompson, Aidan Patrick; Jones, Reese E.; Zimmerman, Jonathan A.; Baskes, Michael I. (University of California, San Diego); Winey, J. Michael (Washington State University); Gupta, Yogendra Mohan (Washington State University); Lane, J. Matthew D.; Ditmire, Todd (University of Texas at Austin); Quevedo, Hernan J. (University of Texas at Austin)

    2011-10-01

    Molecular dynamics simulation (MD) is an invaluable tool for studying problems sensitive to atomscale physics such as structural transitions, discontinuous interfaces, non-equilibrium dynamics, and elastic-plastic deformation. In order to apply this method to modeling of ramp-compression experiments, several challenges must be overcome: accuracy of interatomic potentials, length- and time-scales, and extraction of continuum quantities. We have completed a 3 year LDRD project with the goal of developing molecular dynamics simulation capabilities for modeling the response of materials to ramp compression. The techniques we have developed fall in to three categories (i) molecular dynamics methods (ii) interatomic potentials (iii) calculation of continuum variables. Highlights include the development of an accurate interatomic potential describing shock-melting of Beryllium, a scaling technique for modeling slow ramp compression experiments using fast ramp MD simulations, and a technique for extracting plastic strain from MD simulations. All of these methods have been implemented in Sandia's LAMMPS MD code, ensuring their widespread availability to dynamic materials research at Sandia and elsewhere.

  19. Efficiency promotion for an on-ramp system based on intelligent transportation system information

    Science.gov (United States)

    Xie, Dong-Fan; Gao, Zi-You; Zhao, Xiao-Mei

    2010-08-01

    The effect of cars with intelligent transportation systems (ITSs) on traffic flow near an on-ramp is investigated by car-following simulations. By numerical simulations, the dependences of flux on the inflow rate are investigated for various proportions of cars with ITSs. The phase diagrams as well as the spatiotemporal diagrams are presented to show different traffic flow states on the main road and the on-ramp. The results show that the saturated flux on the main road increases and the free flow region is enlarged with the increase of the proportion of cars with ITS. Interestingly, the congested regions of the main road disappear completely when the proportion is larger than a critical value. Further investigation shows that the capacity of the on-ramp system can be promoted by 13% by using the ITS information, and the saturated flux on the on-ramp can be kept at an appropriate value by adjusting the proportion of cars with ITS.

  20. Lattice-ramp-induced dynamics in an interacting Bose-Bose mixture

    NARCIS (Netherlands)

    J. Wernsdorfer; M. Snoek; W. Hofstetter

    2010-01-01

    We investigate a bosonic quantum gas consisting of two interacting species in an optical lattice at zero and finite temperature. The equilibrium properties and dynamics of this system are obtained by means of the Gutzwiller mean-field method. In particular we model recent experiments where the ramp-

  1. A new proposal to guide velocity and inclination in the ramp protocol for the treadmill ergometer.

    Science.gov (United States)

    Barbosa e Silva, Odwaldo; Sobral Filho, Dário C

    2003-07-01

    To suggest criteria to guide protocol prescription in ramp treadmill testing, according to sex and age, based on velocity, inclination, and max VO2 reached by the population studied. Prospective study describing heart rate (HR), time, velocity, inclination, and VO2 estimated at maximum effort of 1840 individuals from 4 to 79 years old, who performed a treadmill test (TT) according to the ramp protocol. A paired Student t test was used to assess the difference between predicted and reached max VO2, calculated according to the formulas of the "American College of Sports Medicine". Submaximal HR was surpassed in 90.1% of the examinations, with a mean time of 10.0 2.0 minute. Initial and peak inclination velocity of the exercise and max VO2 were inversely proportional to age and were greater in male patients. Predicted Max VO2 was significantly lower than that reached in all patients, except for female children and adolescents (age VO2 actually reached, as a criterion in prescribing the ramp protocol may help in the performance of exercise in treadmill testing. The ramp protocol was well accepted in all age groups and sexes with exercise time within the programmed 8 to 12 minutes.

  2. A new proposal to guide velocity and inclination in the ramp protocol for the treadmill ergometer

    Directory of Open Access Journals (Sweden)

    Odwaldo Barbosa e Silva

    2003-07-01

    Full Text Available OBJECTIVE: To suggest criteria to guide protocol prescription in ramp treadmill testing, according to sex and age, based on velocity, inclination, and max VO2 reached by the population studied. METHODS: Prospective study describing heart rate (HR, time, velocity, inclination, and VO2 estimated at maximum effort of 1840 individuals from 4 to 79 years old, who performed a treadmill test (TT according to the ramp protocol. A paired Student t test was used to assess the difference between predicted and reached max VO2, calculated according to the formulas of the "American College of Sports Medicine". RESULTS: Submaximal HR was surpassed in 90.1% of the examinations, with a mean time of 10.0±2.0 minute. Initial and peak inclination velocity of the exercise and max VO2 were inversely proportional to age and were greater in male patients. Predicted Max VO2 was significantly lower than that reached in all patients, except for female children and adolescents (age < 20 years old. CONCLUSION: Use of velocity, inclination, and maximum VO2 actually reached, as a criterion in prescribing the ramp protocol may help in the performance of exercise in treadmill testing. The ramp protocol was well accepted in all age groups and sexes with exercise time within the programmed 8 to 12 minutes.

  3. The Archival Appraisal of Moving Images: A RAMP Study with Guidelines.

    Science.gov (United States)

    Kula, Sam

    Produced as part of the United Nations Educational, Scientific, and Cultural Organization (UNESCO) Records and Archives Management Programme (RAMP), this publication provides government and non-government archivists and records managers with a comparative study of past and present policies and practices for selecting moving images for…

  4. Integrated Variable Speed Limits Control and Ramp Metering for Bottleneck Regions on Freeway

    Directory of Open Access Journals (Sweden)

    Ming-hui Ma

    2015-01-01

    Full Text Available To enhance the efficiency of the existing freeway system and therefore to mitigate traffic congestion and related problems on the freeway mainline lane-drop bottleneck region, the advanced strategy for bottleneck control is essential. This paper proposes a method that integrates variable speed limits and ramp metering for freeway bottleneck region control to relieve the chaos in bottleneck region. To this end, based on the analyses of spatial-temporal patterns of traffic flow, a macroscopic traffic flow model is extended to describe the traffic flow operating characteristic by considering the impacts of variable speed limits in mainstream bottleneck region. In addition, to achieve the goal of balancing the priority of the vehicles on mainline and on-ramp, increasing capacity, and reducing travel delay on bottleneck region, an improved control model, as well as an advanced control strategy that integrates variable speed limits and ramp metering, is developed. The proposed method is tested in simulation for a real freeway infrastructure feed and calibrates real traffic variables. The results demonstrate that the proposed method can substantially improve the traffic flow efficiency of mainline and on-ramp and enhance the quality of traffic flow at the investigated freeway mainline bottleneck.

  5. Three dimensional experimental investigation of a hypersonic double-ramp flow

    NARCIS (Netherlands)

    Schrijer, F.F.J.; Caljouw, R.; Scarano, F.; Van Oudheusden, B.W.

    The flow over a 15◦-45◦ double compression ramp was studied at Mach 7.5. CFD computations are compared to 2 component PIV (particle image velocimetry) measurements. Furthermore stereoscopic PIV was used to measure the three component velocity vector, enabling to perform a 3D flow survey. The overall

  6. The steep ramp test in healthy children and adolescents: reliability and validity

    NARCIS (Netherlands)

    Bongers, B.C.; Vries, S.I. de; Helders, P.J.M.; Takken, T.

    2013-01-01

    PURPOSE: This study aimed to examine the reliability and validity of the steep ramp test (SRT), a feasible, maximal exercise test on a cycle ergometer that does not require the use of respiratory gas analysis, in healthy children and adolescents.METHODS: Seventy-five children were randomly divided i

  7. Geotechnical characterization of the North Ramp of the Exploratory Studies Facility: Yucca Mountain Site Characterization Project. Volume 1, Data summary

    Energy Technology Data Exchange (ETDEWEB)

    Brechtel, C.E.; Lin, Ming; Martin, E. [Agapito Associates, Inc., Grand Junction, CO (United States); Kessel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

    1995-05-01

    This report presents the results of geological and geotechnical characterization of the Miocene volcanic tuff rocks of the Timber Mountain and Paintbrush groups that the tunnel boring machine will encounter during excavation of the Exploratory Studies Facility (ESF) North Ramp. The is being constructed by the DOE as part of the Yucca Mountain Project site characterization activities. The purpose of these activities is to evaluate the feasibility of locating a potential high-level nuclear waste repository on lands adjacent to the Nevada Test Site, Nye County, Nevada. This report was prepared as part of the Soil and Rock Properties Studies in accordance with the 8.3.1.14.2 Study Plan. This report is volume 1 of the data summary.

  8. Feature and duration of metre-scale sequences in a storm-dominated carbonate ramp setting (Kimmeridgian, northeastern Spain)

    Science.gov (United States)

    Colombié, C.; Bádenas, B.; Aurell, M.; Götz, A. E.; Bertholon, S.; Boussaha, M.

    2014-10-01

    Metre-scale sequences may result from the combined effects of allocyclic and autocyclic processes which are closely inter-related. The carbonate ramp that developed northwest of the Iberian Basin during the late Kimmeridgian was affected by northwestward migrating cyclones. Marl-limestone alternations that settled in mid-ramp environments contain abundant mm to cm thick coarse-grained accumulations that have been related to these events. The aim of this paper is to determine the impact of storm-induced processes on the metre-scale sequence features. Four sections (R3, R4, R6, and R7), which are 5 to 7 m in thickness, were studied bed-by-bed along a 4 km-long outcrop, which shows the transition between the shallow and the relatively deep realms of the middle ramp. Metre-scale sequences were defined and correlated along this outcrop according to the detailed microfacies analysis of host, fine-grained deposits, palynofacies and sequence-stratigraphic analyses, and carbon- and oxygen-isotope measurements. The evolution through time of sedimentary features such as the size of quartz grains and the relative abundance of grains other than quartz (i.e., muscovite, bivalve, ooid, and intraclast) does not correlate from one section to the other, suggesting that the finest as well as the coarsest sediment was reworked in these storm-dominated environments. Small- and medium-scale sequences are defined according to changes in alternation, marly interbed, and limestone bed thickness, and correlated from one section to the other. Because of the effects of storms on sediment distribution and preservation, sequence boundaries coincide with thin alternations and marly interbeds in the most proximal sections (i.e., R3, R4), while they correspond to thin alternations and limestone beds in the most distal sections (i.e., R6, R7). Field observations and palynofacies analyses confirm this sequence-stratigraphic analysis. The excursions in carbon- and oxygen-isotope values are consistent

  9. Double-ramp on the Main Himalayan Thrust revealed by broadband waveform modeling of the 2015 Gorkha earthquake sequence

    Science.gov (United States)

    Wang, Xin; Wei, Shengji; Wu, Wenbo

    2017-09-01

    The 2015Mw 7.8 Gorkha earthquake sequence that unzipped the lower edge of the Main Himalayan Thrust (MHT) in central Nepal provides an exceptional opportunity to understand the fault geometry in this region. However, the limited number of focal mechanisms and the poor horizontal locations and depths of earthquakes in the global catalog impede us from clearly imaging the ruptured MHT. In this study, we generalized the Amplitude Amplification Factor (AAF) method to teleseismic distance that allows us to model the teleseismic P-waves up to 1.5 Hz. We used well-constrained medium-sized earthquakes to establish AAF corrections for teleseismic stations that were later used to invert the high-frequency waveforms of other nearby events. This new approach enables us to invert the focal mechanisms of some early aftershocks, which is challenging by using other long-period methods. With this method, we obtained 12 focal mechanisms more than that in the GCMT catalog. We also modeled the high-frequency teleseismic P-waves and the surface reflection phases (pP and sP) to precisely constrain the depths of the earthquakes. Our results indicate that the uncertainty of the depth estimation is as small as 1-2 km. Finally, we refined the horizontal locations of these aftershocks using carefully hand-picked arrivals. The refined aftershock mechanisms and locations delineate a clear double-ramp geometry of the MHT, with an almost flat décollement sandwiched in between. The flat (dip ∼7 degrees) portion of the MHT is consistent with the coseismic rupture of the mainshock, which has a well-constrained slip distribution. The fault morphology suggests that the ramps, both along the up-dip and down-dip directions, play a significant role in stopping the rupture of the 2015 Gorkha earthquake. Our method can be applied to general subduction zone earthquakes and fault geometry studies.

  10. Flowfield structure and mixing performance of a cantilevered ramp injector%悬臂斜坡喷注器流场结构与混合特性

    Institute of Scientific and Technical Information of China (English)

    毕东恒; 罗世彬; 林志勇

    2015-01-01

    In order to investigate the precise flowfield structure and mixing⁃enhanced mechanism of H2/air mixture for the fore⁃body/inlet of shock⁃induced combustion ramjet( shcramjet) at high Mach number,numerical simulations were conducted to study the H2/air mixing enhancement and flowfield of the cantilevered ramp injector. An implicit finite volume method was used to solve the 3D compressible Navier⁃Stokes equations. Simulation results show that,in general,due to the effect of the ramp,physical phe⁃nomenon,like shock, expansion wave, and longitude vortices etc. are produced in the flowfield. Oblique shock occurs when the freestream meets the ramp. And the flow expands at the edge of the ramp. Longitude vortices appear along with pressure difference, which whirl the air in the movement downstream. So in this way the H2 and air mix better,but the ramp increases the total pressure losses.%为了探索高马赫数下激波诱燃冲压发动机前体/进气道燃料/空气混合的精细流场结构和混合增强机理,采用隐式方法,对悬臂斜坡喷注器进行了三维RANS仿真,得到了喷注器流场的精细结构。仿真结果表明,由于斜坡的作用,流场中产生了激波、膨胀波、流向旋涡等现象。气流经过斜坡时产生了斜激波,并在斜坡边缘处发生膨胀;斜坡侧壁附近在压差的作用下产生了流向旋涡,流向涡在向下游发展过程中卷吸空气,从而增强了混合,但斜坡的存在加大了流场的总压损失。

  11. Large Core Code Evaluation Working Group Benchmark Problem Four: neutronics and burnup analysis of a large heterogeneous fast reactor. Part 1. Analysis of benchmark results. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Protsik, R.; Lewellen, J.W. (eds.)

    1984-01-01

    The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the techniques for modeling the three-dimensional benchmark geometry, and sensitivity studies were carried out to determine the performance parameter sensitivities to changes in the neutronics and burnup specifications. The results of the Benchmark Four calculations indicated that a linked RZ-XY (Hex) two-dimensional representation of the benchmark model geometry can be used to predict mass balance data, power distributions, regionwise fuel exposure data and burnup reactivities with good accuracy when compared with the results of direct three-dimensional computations. Most of the small differences in the results of the benchmark analyses by the different participants were attributed to ambiguities in carrying out the regionwise flux renormalization calculations throughout the burnup step.

  12. 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)

  13. 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.

  14. Mechanomyographic amplitude and mean power frequency responses during isometric ramp vs. step muscle actions.

    Science.gov (United States)

    Ryan, Eric D; Beck, Travis W; Herda, Trent J; Hartman, Michael J; Stout, Jeffrey R; Housh, Terry J; Cramer, Joel T

    2008-03-15

    The purpose of the present study was to compare the mechanomyographic amplitude (MMG(RMS)) and mean power frequency (MMG(MPF)) vs. torque relationships during isometric ramp and step muscle actions for the vastus lateralis (VL) and rectus femoris (RF) muscles. Nineteen subjects (mean+/-S.D. age=24+/-4 years) performed 2 isometric maximal voluntary contractions (MVCs) before and after 2 or 3 isometric ramp muscle actions from (5-95% MVC) to 9 submaximal step muscle actions (15, 25, 35, 45, 55, 65, 75, 85, and 95% MVC). MMG signals were recorded from the VL and RF muscles, and MMG(RMS) and MMG(MPF) values were computed for each corresponding percentage of the MVC. Absolute and normalized MMG(RMS) and MMG(MPF) vs. torque relationships were analyzed and interpreted on a subject-by-subject and composite pattern basis using polynomial regression and repeated measures ANOVAs. For MMG(RMS) and MMG(MPF), only 16-53% and 11-26% of the individual responses were consistent with the composite polynomial models, respectively. In addition, the normalized composite MMG(RMS) values were greater for the RF than the VL from 35 to 85% MVC. Only 47% of the MMG(RMS) and 5% of the MMG(MPF) individual patterns of responses were the same for the ramp and step muscle actions, and differences were also observed for the composite MMG(RMS) and MMG(MPF) patterns between the ramp and step muscle actions. Overall, these findings indicated that the torque-related patterns of responses for MMG(RMS) and MMG(MPF) were different among subjects (i.e., inter-individual variability) and were muscle- (VL vs. RF) and mode-specific (ramp vs. step).

  15. Roles of CLR/RAMP Receptor Signaling in Reproduction and Development

    Science.gov (United States)

    Chang, Chia Lin; Hsu, Sheau Yu Teddy

    2016-01-01

    Adrenomedullin (ADM), calcitonin gene-related peptides (α- and β-CGRPs), and intermedin/adrenomedullin 2 (IMD/ADM2) are major regulators of vascular tone and cardiovascular development in vertebrates. Recent research into their functions in reproduction has illuminated the role of these peptides and their cognate receptors (calcitonin receptor-like receptor/receptor activity-modifying protein (CLR/RAMP) receptors) in fetal–maternal blood circulation, feto-placental development, female gamete development, and gamete movement in the oviduct. Although ADM family peptides function in a temporally and spatially specific manner in various reproductive processes, they appear to act via a similar set of second messengers, including nitric oxide, cyclic GMP, cyclic AMP, and calcium-activated potassium channels in different tissues. These discoveries supported the view that CLR/RAMP receptors were recruited to perform a variety of newly evolved reproductive functions during the evolution of internal reproduction in mammals. These advances also provided insight into how CLR/RAMP receptor signaling pathways coordinate with other physiological adaptions to accommodate the extra metabolic needs during pregnancy, and captured some important details as to how fetal–maternal vascular communications are generated in the first place. Furthermore, these findings have revealed novel, promising opportunities for the prevention and treatment of aberrant pregnancies such as pregnancy-induced hypertension, preeclampsia, and tubal ectopic pregnancy. However, significant efforts are still needed to clarify the relationships between certain components of the CLR/RAMP signaling pathway and aberrant pregnancies before CLR/RAMP receptors can become targets for clinical management. With this understanding, this review summarizes recent progresses with particular focus on clinical implications. PMID:23745703

  16. Blank corrections for ramped pyrolysis radiocarbon dating of sedimentary and soil organic carbon.

    Science.gov (United States)

    Fernandez, Alvaro; Santos, Guaciara M; Williams, Elizabeth K; Pendergraft, Matthew A; Vetter, Lael; Rosenheim, Brad E

    2014-12-16

    Ramped pyrolysis (RP) targets distinct components of soil and sedimentary organic carbon based on their thermochemical stabilities and allows the determination of the full spectrum of radiocarbon ((14)C) ages present in a soil or sediment sample. Extending the method into realms where more precise ages are needed or where smaller samples need to be measured involves better understanding of the blank contamination associated with the method. Here, we use a compiled data set of RP measurements of samples of known age to evaluate the mass of the carbon blank and its associated (14)C signature, and to assess the performance of the RP system. We estimate blank contamination during RP using two methods, the modern-dead and the isotope dilution method. Our results indicate that during one complete RP run samples are contaminated by 8.8 ± 4.4 μg (time-dependent) of modern carbon (MC, fM ∼ 1) and 4.1 ± 5.5 μg (time-independent) of dead carbon (DC, fM ∼ 0). We find that the modern-dead method provides more accurate estimates of uncertainties in blank contamination; therefore, the isotope dilution method should be used with caution when the variability of the blank is high. Additionally, we show that RP can routinely produce accurate (14)C dates with precisions ∼100 (14)C years for materials deposited in the last 10,000 years and ∼300 (14)C years for carbon with (14)C ages of up to 20,000 years.

  17. 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.

  18. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2014-09-01

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature can be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.

  19. Sedimentology and preservation of aeolian sediments on steep terrains: Incipient sand ramps on the Atacama coast (northern Chile)

    Science.gov (United States)

    Ventra, Dario; Rodríguez-López, Juan Pedro; de Boer, Poppe L.

    2017-05-01

    The origin of topographically controlled aeolian landforms in high-relief settings is difficult to synthesize under general models, given the dependence of such accumulations on local morphology. Quaternary sand ramps have been linked to palaeoclimate, regional geomorphology and wind patterns; however, controls on the early development and preservation of such landforms are poorly known. This study describes the morphology and sedimentology of complex sedimentary aprons along steep coastal slopes in the Atacama Desert (Chile). Direct slope accessibility and continuous stratigraphic exposures enable comparisons between active processes and stratigraphic signatures. Stratigraphic facies distribution and its links with patterns of aeolian deposition show that the preservation of wind-laid sediments depends on the morphology and processes of specific slope sectors. The spatial organization of runoff depends on bedrock configuration and directly controls the permanence or erosion of aeolian sediment. The occurrence of either water or mass flows depends on the role of aeolian fines in the rheology of flash floods. In turn, the establishment of a rugged surface topography controlled by patterns of mass-flow deposition creates local accommodation for aeolian fines, sustaining the initial aggradation of a colluvial-aeolian system. By contrast, slopes subject to runoff develop a thin, extensive aeolian mantle whose featureless surface is subject mostly to sediment bypass down- and across-slope; the corresponding stratigraphic record comprises almost exclusively thin debris-flow and sheetflood deposits. Slope morphology and processes are fundamental in promoting or inhibiting aeolian aggradation in mountain settings. Long-term sand-ramp construction depends on climate and regional topography, but the initial development is probably controlled by local geomorphic factors. The observed interactions between wind and topography in the study area may also represent a process

  20. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  1. Determining initial enrichment, burnup, and cooling time of pressurized-water-reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    Science.gov (United States)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/137Cs, 134Cs/137Cs, 106Ru/137Cs, and 144Ce/137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity's behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

  2. Overseas Tech Firms Ramp Up Hiring In Silicon Valley

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    @@ Fpr years, U.S.high-tech executives have fretted about the loss of technology jobs to countries such as China and India.But there is a countervailing force: Foreign companies are leaning heavily on engineers they hire in Silicon Valley. Companies from Europe and Asia increasingly are using high-tech teams in the Bay Area to conduct long-range research and help design cutting-edge products for global markets.

  3. Hydrogeology of the unsaturated zone, North Ramp area of the Exploratory Studies Facility, Yucca Mountain, Nevada

    Science.gov (United States)

    Rousseau, Joseph P.; Kwicklis, Edward M.; Gillies, Daniel C.; Rousseau, Joseph P.; Kwicklis, Edward M.; Gillies, Daniel C.

    1999-01-01

    Yucca Mountain, in southern Nevada, is being investigated by the U.S. Department of Energy as a potential site for a repository for high-level radioactive waste. This report documents the results of surface-based geologic, pneumatic, hydrologic, and geochemical studies conducted during 1992 to 1996 by the U.S. Geological Survey in the vicinity of the North Ramp of the Exploratory Studies Facility (ESF) that are pertinent to understanding multiphase fluid flow within the deep unsaturated zone. Detailed stratigraphic and structural characteristics of the study area provided the hydrogeologic framework for these investigations. Multiple lines of evidence indicate that gas flow and liquid flow within the welded tuffs of the unsaturated zone occur primarily through fractures. Fracture densities are highest in the Tiva Canyon welded (TCw) and Topopah Spring welded (TSw) hydrogeologic units. Although fracture density is much lower in the intervening nonwelded and bedded tuffs of the Paintbrush nonwelded hydrogeologic unit (PTn), pneumatic and aqueous-phase isotopic evidence indicates that substantial secondary permeability is present locally in the PTn, especially in the vicinity of faults. Borehole air-injection tests indicate that bulk air-permeability ranges from 3.5x10-14 to 5.4x10-11 square meters for the welded tuffs and from 1.2x10-13 to 3.0x10-12 square meters for the non welded and bedded tuffs of the PTn. Analyses of in-situ pneumatic-pressure data from monitored boreholes produced estimates of bulk permeability that were comparable to those determined from the air-injection tests. In many cases, both sets of estimates are two to three orders of magnitude larger than estimates based on laboratory analyses of unfractured core samples. The in-situ pneumatic-pressure records also indicate that the unsaturated-zone pneumatic system consists of four subsystems that coincide with the four major hydrogeologic units of the unsaturated zone at Yucca Mountain. In

  4. Most na rampě křižovatky v Brně

    OpenAIRE

    Novotný, Jan

    2013-01-01

    Práce se zabývá řešením půdorysně zakřiveného mostu na rampě v Brně. Navrženou nosnou konstrukcí je spojitý nosník tvořený dvoukomorou. Na tuto konstrukci jsou uvažovány účinky dopravy dle normy ČSN EN 1991-2. Na základě výsledných namáhání je konstrukce dimenzována. Thesis deals with horizontally curved bridge on the ramp in Brno. Designed structure is continous beam formed by bicameral shape, on which are considered the effects of traffic according to ČSN EN 1991-2. Based on the results ...

  5. Validity of Thermal Ramping Assays Used to Assess Thermal Tolerance in Arthropods

    DEFF Research Database (Denmark)

    Overgaard, Johannes; Kristensen, Torsten Nygård; Sørensen, Jesper Givskov

    2012-01-01

    Proper assessment of environmental resistance of animals is critical for the ability of researchers to understand how variation in environmental conditions influence population and species abundance. This is also the case for studies of upper thermal limits in insects, where researchers studying...... animals under laboratory conditions must select appropriate methodology on which conclusions can be drawn. Ideally these methods should precisely estimate the trait of interest and also be biological meaningful. In an attempt to develop such tests it has been proposed that thermal ramping assays...... empirically test these model predictions using two sets of independent experiments. We clearly demonstrate that results from ramping assays of small insects (Drosophila melanogaster) are not compromised by starvation- or dehydration-stress. Firstly we show that the mild disturbance of water and energy balance...

  6. Impacts of ramping inflexibility of conventional generators on strategic operation of energy storage facilities

    DEFF Research Database (Denmark)

    Nasrolahpour, Ehsan; Kazempour, Jalal; Zareipour, Hamidreza

    2016-01-01

    This paper proposes an approach to assist a pricemaker merchant energy storage facility in making its optimal operation decisions. The facility operates in a pool-based electricity market, where the ramping capability of other resources is limited. Also, wind power resources exist in the system....... The merchant facility seeks to maximize its profit through strategic inter-temporal arbitrage decisions, when taking advantage of those ramp limitations. The market operator, on the other hand, aims at maximizing the social welfare under wind power generation uncertainty. Thus, a stochastic bi......-level optimization model is proposed, taking into account the interactions between the storage facility and the market operator, and the existing market opportunities for the storage facility. The proposed bilevel model is then transformed into a Mathematical Program with Equilibrium Constraints (MPEC) that can...

  7. The Tool Life of Ball Nose end Mill Depending on the Different Types of Ramping

    Directory of Open Access Journals (Sweden)

    Vopát Tomáš

    2014-12-01

    Full Text Available The article deals with the cutting tool wear measurement process and tool life of ball nose end mill depending on upward ramping and downward ramping. The aim was to determine and compare the wear (tool life of ball nose end mill for different types of copy milling operations, as well as to specify particular steps of the measurement process. In addition, we examined and observed cutter contact areas of ball nose end mill with machined material. For tool life test, DMG DMU 85 monoBLOCK 5-axis CNC milling machine was used. In the experiment, cutting speed, feed rate, axial depth of cut and radial depth of cut were not changed. The cutting tool wear was measured on Zoller Genius 3s universal measuring machine. The results show different tool life of ball nose end mills depending on the copy milling strategy.

  8. Predictive Control of Wind Turbine for Load Reduction during Ramping Events

    DEFF Research Database (Denmark)

    Liu, Weipeng; Li, Changgang; Liu, Yutian

    2017-01-01

    With increasing penetration of wind power, the impact of its intermittence and volatility on power systems becomes more severe. A predictive control strategy for wind turbines (WTs) is proposed to deal with wind power ramping events and reduce WT load on the blades. The blade load model is based...... on the Blade Element Momentum (BEM) theory. The generator speed and pitch angle are simultaneously regulated to realize the control objectives. A two-stage optimization is designed in order to reduce the computational complexity. The objectives of the first stage are minimizing the ramping rate and maximizing...... the power generation. A trade-off is made between the two contradictory objectives by setting weight coefficients. The second stage reduces the WT load and meanwhile guarantees the power reference from the first stage is tracked. Feedback is designed based on neural network prediction to compensate...

  9. Analysis of Cliff-Ramp Structures in Homogeneous Scalar Turbulence by the Method of Line Segments

    Science.gov (United States)

    Gauding, Michael; Goebbert, Jens Henrik; Peters, Norbert; Hasse, Christian

    2015-11-01

    The local structure of a turbulent scalar field in homogeneous isotropic turbulence is analyzed by direct numerical simulations (DNS). A novel signal decomposition approach is introduced where the signal of the scalar along a straight line is partitioned into segments based on the local extremal points of the scalar field. These segments are then parameterized by the distance between adjacent extremal points and a segment-based gradient. Joint statistics of the length and the segment-based gradient provide novel understanding about the local structure of the turbulent field and particularly about cliff-ramp-like structures. Ramp-like structures are unveiled by the asymmetry of joint distribution functions. Cliff-like structures are further analyzed by conditional statistics and it is shown from DNS that the width of cliffs scales with the Kolmogorov length scale.

  10. From Finding Aids to Wiki Pages: Remixing Archival Metadata with RAMP

    Directory of Open Access Journals (Sweden)

    David González

    2013-10-01

    Full Text Available The Remixing Archival Metadata Project (RAMP is a lightweight web-based editing tool that is intended to let users do two things: (1 generate enhanced authority records for creators of archival collections and (2 publish the content of those records as Wikipedia pages. The RAMP editor can extract biographical and historical data from EAD finding aids to create new authority records for persons, corporate bodies, and families associated with archival and special collections (using the EAC-CPF format. It can then let users enhance those records with additional data from sources like VIAF and WorldCat Identities. Finally, it can transform those records into wiki markup so that users can edit them directly, merge them with any existing Wikipedia pages, and publish them to Wikipedia through its API.

  11. Versatile CAMAC power supply controller-monitor with built-in ramping and ripple monitor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, H.; Horelick, D.

    1984-10-01

    An integrated power supply controller-monitor has been designed and is in use to control large power supplies for SLC dc magnets. This single-width CAMAC module contains a 14-bit DAC, a 14-bit ADC, and several channels of optically coupled digital status and control signals. Additional features include built-in selectable ramping rates, self-test capabilities, and a ripple monitor circuit to measure ac ripple in the power supply current.

  12. Validity of thermal ramping assays used to assess thermal tolerance in arthropods.

    Science.gov (United States)

    Overgaard, Johannes; Kristensen, Torsten Nygaard; Sørensen, Jesper Givskov

    2012-01-01

    Proper assessment of environmental resistance of animals is critical for the ability of researchers to understand how variation in environmental conditions influence population and species abundance. This is also the case for studies of upper thermal limits in insects, where researchers studying animals under laboratory conditions must select appropriate methodology on which conclusions can be drawn. Ideally these methods should precisely estimate the trait of interest and also be biological meaningful. In an attempt to develop such tests it has been proposed that thermal ramping assays are useful assays for small insects because they incorporate an ecologically relevant gradual temperature change. However, recent model-based papers have suggested that estimates of thermal resistance may be strongly confounded by simultaneous starvation and dehydration stress. In the present study we empirically test these model predictions using two sets of independent experiments. We clearly demonstrate that results from ramping assays of small insects (Drosophila melanogaster) are not compromised by starvation- or dehydration-stress. Firstly we show that the mild disturbance of water and energy balance of D. melanogaster experienced during the ramping tests does not confound heat tolerance estimates. Secondly we show that flies pre-exposed to starvation and dehydration have "normal" heat tolerance and that resistance to heat stress is independent of the energetic and water status of the flies. On the basis of our results we discuss the assumptions used in recent model papers and present arguments as to why the ramping assay is both a valid and ecologically relevant way to measure thermal resistance in insects.

  13. Validity of thermal ramping assays used to assess thermal tolerance in arthropods.

    Directory of Open Access Journals (Sweden)

    Johannes Overgaard

    Full Text Available Proper assessment of environmental resistance of animals is critical for the ability of researchers to understand how variation in environmental conditions influence population and species abundance. This is also the case for studies of upper thermal limits in insects, where researchers studying animals under laboratory conditions must select appropriate methodology on which conclusions can be drawn. Ideally these methods should precisely estimate the trait of interest and also be biological meaningful. In an attempt to develop such tests it has been proposed that thermal ramping assays are useful assays for small insects because they incorporate an ecologically relevant gradual temperature change. However, recent model-based papers have suggested that estimates of thermal resistance may be strongly confounded by simultaneous starvation and dehydration stress. In the present study we empirically test these model predictions using two sets of independent experiments. We clearly demonstrate that results from ramping assays of small insects (Drosophila melanogaster are not compromised by starvation- or dehydration-stress. Firstly we show that the mild disturbance of water and energy balance of D. melanogaster experienced during the ramping tests does not confound heat tolerance estimates. Secondly we show that flies pre-exposed to starvation and dehydration have "normal" heat tolerance and that resistance to heat stress is independent of the energetic and water status of the flies. On the basis of our results we discuss the assumptions used in recent model papers and present arguments as to why the ramping assay is both a valid and ecologically relevant way to measure thermal resistance in insects.

  14. Heart Rate Variability as a Measure of Airport Ramp-Traffic Controllers Workload

    Science.gov (United States)

    Hayashi, Miwa; Dulchinos, Victoria Lee

    2016-01-01

    Heart Rate Variability (HRV) has been reported to reflect the person's cognitive and emotional stress levels, and may offer an objective measure of human-operator's workload levels, which are recorded continuously and unobtrusively to the task performance. The present paper compares the HRV data collected during a human-in-the-loop simulation of airport ramp-traffic control operations with the controller participants' own verbal self-reporting ratings of their workload.

  15. An EOQ model with time dependent Weibull deterioration and ramp type demand ,

    Directory of Open Access Journals (Sweden)

    Chaitanya Kumar Tripathy

    2011-04-01

    Full Text Available This paper presents an order level inventory system with time dependent Weibull deterioration and ramp type demand rate where production and demand are time dependent. The proposed model of this paper considers economic order quantity under two different